ML20005H150: Difference between revisions
StriderTol (talk | contribs) (StriderTol Bot change) |
StriderTol (talk | contribs) (StriderTol Bot change) |
||
| Line 1,637: | Line 1,637: | ||
3 E 5* | 3 E 5* | ||
#d>O@' IMAGE EVALUATION | #d>O@' IMAGE EVALUATION ((j/ 8/4 F | ||
4x restroest wrm pppp p | 4x restroest wrm pppp p | ||
I.0 E 2 EM 5!NE l 1 | I.0 E 2 EM 5!NE l 1 | ||
Latest revision as of 14:37, 15 March 2020
| ML20005H150 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/12/1990 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20005H144 | List: |
| References | |
| NUDOCS 9001240207 | |
| Download: ML20005H150 (165) | |
Text
{{#Wiki_filter:. - . - - F ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE , SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 i DOCKET NOS. 50-327 AND 50-326 I
.- .. i t (TVA-SQN-TS-89-33) i g . .,
I< f LIST OF AFFECTED PAGES I Unit 1 i
! y i VII . i B 2-1 B 2-3 B 2-5 l 3/4 1-19 3/4 2-10 3/4 2-11
. 3/4 2-12 '
3/4 2-13 ' 3/4 2-14
+
3/4 2-18 . jf 3/4 2-19
- " B3/4 2-1 B3/4 2-2 .
i' B3/4 2-4 # 9 B3/4 2-5 B3/4 4-1 Unit 2 , l0 y : VII B 2-1 B 2-3 5 2-5 - 3/4 1-19
- p. g 3/4 2-8 -
t "
. . . 3/4 2-9
. 3/4 2-10 ^
3/4 2-11 3/4 2-12 3/4 2-16 ' 3/4 2-17 B3/4 2-1 B3/4 2-2 B3/4 2-4 9001240207 90011, B3/4 2-5 PDR ADOCK 05000327 B3/4 4-1 P , PDC ,
em l l INDEX I LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQ _, i
. .=.:.-
IECTION PAGE
.3/4.2 . POWEf! DISTRIBUTION LIMITS 3/4.2.1 Axial Flux Difference..................................... 3/42-1 3/4.2.2 Heat Flux Hot Channel Factor....................
3/4 2-5 # 3/4.2,'J c/S R floyfat/ M@ Nue r,.e e.t EinnecPy.Nor du. . . . . . . . sA W e t hecTAR -I
. ...................................... 3/4 2-10 AR l 3/4.2.4 Quadrant Power Tilt Ratio................................. , 3/4 2-15 t
3/4.2.5 DNB Parameters.........i.................................. 3/4 2-18 3 3/4.3 INSTRUMENTATION 3/4.3.1 ; REACTOR TRIP SYSTEM INSTRUMENTATION....................... 3/4 3-1 3/4.3.2 . ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION.........................................' 3/4 3-14 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation...................... 3/4 3-39 i Movable Incore Detectors.................................. 3/4 3-43 Seismic Instrumentat(on................................... 3/4 3-44 ; Meteorological Instrumentation............................ 3/4 3-47 Remote Shutdown Instrumentation........................... 3/4 3-50 Chlorine Detection Systems (Deleted)
' 3/4 3-54 R66 Accident Monitoring Instrumentation....................... 3/4 3-55, Fire Detection Instrumentation............................ 3/4 3-58 Radioactive Liquid Effluent Monitoring Instrumentation.... F 3/4 3-69 Radioactive Gaseoas' Effluent Monitoring Instrumentation...
3/4 3-74 SEQUDYAH - UNIl~ 1 V
, Amendment No. 62 October 30, 1987
T: \ L j i
+
I' C' . . . , - INDEX
%{
j BASES , i
~~~~"__
t u -
\
4 't-i SECTION . i
. . . . . . . . . _PAGE j 3/4.0
- APPLICABILITY................................................
, B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS '
3/4.1.1 BORATION C0HTROL...........................................
, B 3/4 1-1 i 3/4.1.2 BORATION SYSTEMS........................................... B 3/4~1-2 1
3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................. : B 3/4 1-3 - 3/4.2 POWER DISTRIBUTION LIMITS , F 3/4.2.1 AXIAL FLUX DIFFEP.ENCE...................................... B 3/4 2-1 g 1/4.2.2 and 3/4.2.3 HEAT FLUX ... . .. - .. AND NUCLEAR ENTHALPY N HOT CHANNELJ=..:.F g G0.0.^!I.yp B3 /4 2-2
~
( 3/4.2.4 QUADRANT POWER TILT 3/4.2.5 DNB RATI0.................................. B 3/4 2-5 PARAMETERS............................................. B 3/4 2-5 Y,i 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION............ . .. .... ... ......... B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION. ................ B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION................................. B 3/4 3-2
- 3/4.4 REACTOR COOLANT SYSTEM .
TP 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION.............. B 3/4 4-1 3/4.4.2 and 3/4.4.3 SAFETY AND RELIEF VALVES....................... B 3/4 4-l
- 3/4.4.4 '
PRESSUR12ER................................................ B 3/4 4-2
- 3/4.4.5 STEAM GENERATORS .......................................... B 3/4 4-2 ,
e r- l SEQUOYAH - UNIT 1 XII W D 1980
. _ . - _ . _ _ _ . . - - ~__ . . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ .
- i
** * )
(' l g,4 5AffTV LIMITS L L BASE 5 . l L ~
-l tarmjatm aird W 5 a Wg%mutsan L
% es9 egredgn w l n 4 \
t The restrictions of thii safety limit prov'est everheating of the fuel and I i products cladding possible ta the reacter perforatten eselant.which umuld result la the release of fission j ; Overheating of the fuel eladding is prevented ,g-ay vestricting fuel opprettes to within the nucleate boiling regime where the * ' heat transfer coefficient is large and the cladding surface temperature is I. ; alightly above the coolant asturation tamperature. " 3-Operation above the upper boundary of the nucleate boiling regime could vesult in escessive claddtag temperatures because of the enset of departure free nucleate 64111ag (DNB) and the resultant sharp reduction in heat transfer k i i esefficient. OS is not a directly measurable parameter during aperation and 3 therefore T'ERMA. POWER ar4 cter Coelant Temp to and presnure have_toen related te KNB mrough cerfew" The developed to predict the lux and the locati DNBice M A @ly unifera 'beeh-DN5 for an'a and non uniform heat flux distributions. ; definedastheratieoftheheatflunthatwouldcauseDNSataparticularThe local - core location to the local heat flux, is indicative of the asegin to DNB. y M get ' The niinimum va - aper oftheDNBRdurir'Tran anticipated atty state opeion, normal j, value ce nel transients,95 ponds to a ent probability a ts is limite 1.30. This . C that'0NB wi et ecsur and is esen as an appropr 95 percent con margin to DN nee level r all operatino cond leris. The curves of Figure 2.1-1 show the 1eci of points of THERMAL
- p0WER Roscter Coolant Systee_a m sure and everage temperature for which the M18l 23, or the avermi enthalev at the vessel emit minimum is equal to theDNBR enthalpy is ne lessliould, of saturated tha[nthe safety analysis DNBR limit ;
The curves are based on an enthalpy het channel factorgF" , of 1.55 and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase ing F" at reduced power based on the empression: Uli F," = 1.55 [1+ 0.3 (1-P)) 4118j where p is the fraction of RATED THERMAL POWER
- 9
( . SEQUDYAN - L8 TIT 1 4 2-1 Amendment No. 10. 114 May 3.1989 e a n,r
. _ _ _ __ _ _ _ . _ - . . _ _ _ . _ _ . _ _ . . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ - . _ . . .._ ..m._,
t l h g.:.._ . - .
=
. . y-. - _ .._ . _ ,,,
.. ._ y- ._,
- 7 18 ERT 1_ m.
. - . . . . . . . . - _m --
4i . . __ _ : - -- -.. . . _ . . . The DNB.
' percent probabilitydesign basis..is.at that th followst' there'aust-be at least e 95 e minimum DNBR of the limiting rod during Condition I and II events is greater than or egual to the DNBR limit .
this application).of the DNS correlation being used (the WRB-1 or W-3 correlai ' the entire applicable experimental data set such that there is a 95The when the minimum DNBR is at the DNBR limit, percent probability wit i
'.! :'2 ' ' '.. .
m
- r. . .
1 t' . * . ss : l
*T.*=..- * * . . . . . = . . . ,
., s. . l l
. . 1 1
.. . ..w.. ......s ' ' '
4 l 4 s Q l t l 9 .
. ., **P
. - - e
. - - ~ , - , . . , . _ , , . . . _ _ - _ _ . , .
. a i
$AFETY LIMITS ^
l BASES Q Manual Reactor Trio '
- i instrumentation channels and provides manual reactor trip; power Rance. Neutron F10x ~
r - The Power Range. Neutron Flux'chennel high setpoint provides reactor core ~ protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. The low set point provides -
- l redundant protection in the power range for a power excursion beginning from low power.. The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicate a power ,
level of above approximately 10 percent of RATED THERMAL POWER) and is auto. fR; matica11y reinstated when P-10 becomes inactive (three of the four channels indicate a power level below approximately 9 percent of RATED THERMAL POWER). Power Rance. Neutron Flux. High Rates The Power Range Positive Rate trip provides protection against rapid flux ; increases which are characteristic of rod ejection events from any r level. . Specifically, this trip complements the Power Range Neutron Flux Hi h and Low ' trips to ensure that the criteria are met fo u nd ainetian from na ta) _ 7. The Power Range Negative Ra yNovicesprotectiontoensurethatsnethe s
- minimum DNBR is maintained above for control rod drop accidents. At high when in conjunction with nuclear power being maintained r equ power by action of the automatic rod control system, could cause an unconserva- -
tive local DNBR to exist. The Power Range Negative Rate trip will prevent this i from occurring by tripping the reactor for all single or multiple dropped rods. Intermediate and Source Rance. Nuclear Flux , l The Intemediate and source Range, Nuclear Flux trips provide reactor core protection during reactor startv>. . . tion to the low setpoint trip of the Jower Range, Neutron Flux channel r source Range Channels will ' initiate a reactor trip at about 10+5 counts per second unless manually blocked when P-6 becomes active. - The Intermediate ' 5-SEQUOYAH - UNIT 1 8 2-3 Revised 08/18/87 ' .
,. - - - - - , _ . . . - _ _ . - __ _.,,_--,--.e -
.,,.,,m, -
,, , - , - - -- .,,,y.,.--- ,,,,,,m-3,y
f i ( , '5AFETY LIMITS- - o
\
)ASES :
L . i analyses; however, its functional capability at the specified trip setting is required by this specification to enhawe the overall reliability of the Reactor protection System. .; l pressuriter Pressure l i The Pressurizer High and Low p'ressure trips are provided to limit the : pressure range in which reactor operation is permitted. The High pressure trip is backed up by the pressurizer code safety valves for RCS overpressure .! i protection, and is therefore set lower than the set pressure for these valves
.(2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure. ,
Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant 5ystem overpressurization by limiting the water level to a volume l sufficient to retain.a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however its functional capabiltty at the specified i-tripsettingisrequiredbythIs'specificationtoenhancetheoverallreliability of the Reactor Protection System.*
' Loss of-Flow , , , , . . , , , , g . , .
The Loss of Flow trips provide core protection to preved DNB in the event of a loss of one or more reactor coolant pumps. l Anove 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 89% of nominal full loop flow. Above 36% (P-8
. the flow in any) of RATED THERMAL POWER, automatic reactor trip will occur single loa) drops below 895 of nominal full loop flow. "
latter trip will prevent tw minimum value of the DNBR from going below[g.s_ Et during normal operational transients and anticipated transients when 3 loops are in operation and the Overtemperature Delta T trip set point is adjusted to the value specified for all loops in operation. trip set point adjusted to the value specified for 3Withcop theoperation, Overtemperature the P-8 Delta Y trip at 76% THERMAL #0WER will prevent the minimum value of the DNBR from going below during normal operational transients and anticipated transients with 3 loopshin operation. j l k.. the safety analysis DNBR limit SEQUOYAH UNIT 1 - B 2-5 y, e ._,,.e. _ _ . , _ , - . . , . _ . ,,m. , ,_ , _. .._
-. 1
, REACTIVITY CONTROL SYSTEMS-
> C R00 DROP TIME 2.7
=-. - -- : : ~~- .
' ~* '
-- K~= ;
{!MITINGCONDITIONFOROPERATION .. ,_ f 3.1. 3. 4 The indivic'. e' full
- the fully withdrawn psition, length (shutdown and control) ri p time from shall be less than or equal to 1112e '
51 beginning of doct/ of stationary gripper coil voltage to dashpot entry with: ; a. T ,, greater than or equal to $41'F, and b.' All reactor coolant pumps operating. APPLICA81LITY: MDDES 1 and 2 q ACTION: s. With the drop time of any full length rod determined to exceed prior to proceeding to N0DE 1 or 2.the above limit, restore the , b. With the rod drop times within limits but determined with 3 reactor coolant pumps operating, operation may proceed provided THERNAL
- POWER POWER is--restricted to less than or equal to 71% of RATED THERMAL f
., . p 7 g , , , i- - - * * ,
_ SURVEILLANCE REQUIREMENTS .
.....4.1.3.4 n
measurement prior to reactor criticality:The rod drop tierd of full l . a. For all rods following each removal of the reactor vessel head, . b. For specifically affected individual rods following any main- - tenance on or modification to the control rod drive system
. which could affect the drop time of those specific rods, and ,
~
c. At least once pe,r 18 months. ' l .
#Fu,11y withdrawn shall be the condition where shutdown and contiol b P
l at a position within the interval of > 222 and < 231 steps withdrawn, inclusive. ~ ~ all2 L SEQUOYAH - UNIT 1 3/A 1-19 Amendment No. 108 , March 28,1989
--_w. * - ___,-_______m . _ .
. , . . - . , , , - ,, . _ - , . . . w..,, ,__...,,..-_.,m..,.. . . , _ ,
.. 1 l
[ 3/4.L3 #UCLEM ENTHALPY HOT CHAWEL FACTOR- Fj
- I powR etsuteur:0N trurTs 3 2.3. Fj shall be lknPed by h
> Y 4. h th FL M ATE M R bl!Ouisnd;helol06 Ship! '
I
- ...):.;.h::::.'y::~;..:.;, c ; .. . _.. . -
l L!wff!NG CONDITION FOR OPERATION
- 3.3.3 The instion of icated Rea Coolant Sys h (RCs) total * '
ow rata and .R 2shall intained wi h intheregion{efallowable spe tien shown e igure 3.t-3 e 4 1eop ope tion: N !
' '~
wherop
, g 1
, H \ d J. ,
1.43 L1.0 + 0 3 (1.0 - P N , g i L ... . s R. +- Fa *I.55 D.o+,o,.30.o4 El NRBP.(Bu H ,
- Qp = WERMA RATFD TNF EMA' pnWFD 1 NR '
. Fh = Measured lues of g obtained by sing the ,
able _.
~
incere de es to obtain a power d tribution .
~
measured lues of Fg N\ all h be use to calcu1a Rs e Figure 3. -3 include measurement' neertainti of 3. for flow a 4E for i
. . ... .. - _. .. e6seasurenetofFh.a
- e. R8P'(tu = Rod Penalty a function f region a age
. ..bvenup 's shown in Where a reg n '
is deft d as those igure 3.2-4.it.h the same
. semblies w i Ioading $e (reloads) r enrichseht (first co .
APPLICA81LITY: MODE 1 ' g: .... .,
' ~ '
th the comb tion of RCS tal flow ra and Rg , outside t regions e cceptable ration "shown Figues 3. : s Within urs:
- 1. Eithe store the e ination of CS total f rate and Ry , to within above limi or '
N2.Teduce NE
. ,{nggf g, . M R to le than M o RATED N ER R and red the Power e Neutron ux - High ,
tr setpoint less than or ual to 55% RATED THE L POWE within the t 4 hours.
- December 23, 1982 SEQUOYAH - UNIT 1 .
3/4 2-10 Amendment No.19 - O e
- 4
I i (
. .- .- .. . . . ... _ _ . , i
. . . . _ , . , . .~ ... n .. ,
. . .w. .-. _: ..::. . INSERT 2 . . _ - .
1 f . 4
+-
.g.i.a.,
._ .. .c1 . =w"r' : '- ~ n.~1... ..; . . .
With F g exceeding its limits . .- - ' 4
- a. .
Reduce
, , , , ...THERMAL
. . . . . . ~ . , POWER .... to l.ess than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High' Trip '
Satpoints to 5 55 of RATED THERMAL POWER within the next 4 hours, b. Demonstratethruin-coremappingthat(H is within its limit , within'24 hours after exceeding the limit or reduce THERMAL POWER , to less than 5%.of RATED THERMAL POWER within the next 2 hours, C and
- c. Identify and correct the cause of the out of limit condition prior to increasing'TFERNAL POWER above the reduced limit required by. . v a. or b. above; subsequent POWER OPERATION may w..~........~.
proceedprovidedthat(Hisdemonstratedthroughin-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of ' RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours after attaining 95% of greater RATED THERMAL . POWER. - I i l e 4
- . , , , , - . . --.,....r w- ---- --e.-.v-*---*---+- *en.-*-+-w-+*
w--'s* * - - -* * " - += * *+ + - -
$!i- , p POWER DISTRIRUTION LIMITS
. i
' I
- (Conti ). '
i
. . b. .tilthin 24 rs of in tally being ide the ~
rify th ve limits. - incore f1 espping RCS total ow rata ~ rison the the combi ion of Rg , and RCS to 1 flow ..
~
~
rate re reste to within above Its , or reduc { 4 THE POWER to le than SE e RATED POWER with the next hours. * '
- c. ntify and erect the e se of the -of-limit ndition l pri to increas g THERMAL R above t reduced RMAL i p0WER imit requi
' 'by ACTION taas a.2 a or b. above, subseque t POWER,0PE T10N any p i-eed provide that the ' combinati of Rj , R2e Nl indicated C5 total f1 rate are ' L nstrated, through inc flux mapp g and RCS tal flow i i; _ ra comparison, to be with i
- the region of acceptab, s opera on shown.i Figure 3.2- rior to e eeding the '
I '\ L followin ' THERMAL R levels:- '
. A nomi 50E of RA THERMAL R.
- 2. A nominal .* of RATED ERMAL POWE and
- 3. hin 24 hour of attain of greater or equal 95%
TED THERMA POWER. s SURVEILLAN 's REQUIREME x5 ? s s 4.2.'.1 The pro isions of ification .0.4 are applicable. s 4.2.3.2 The combi ion of indi ted RCS to i flow rat and R , R g shall be termined to .e within t regions o cceptable 2 tration
. f Figure 3. 3: .
~
Q * [s' . SEQUOYAN - UNIT 1 . 3/4 2-n e
~ - - , - - - ,-e-~--.--,r----. -- -,~ -.-n-,-- =
l ( -
~ ~ - .
Inser- 5 _ . . . . _ I S0WEILLANCE REQUIRDeffs - 4.1.3.1 The provisions of Specification 4.0.4 are not applicable. . 4.1.3.2 F"gshall be detemined to be within its limit by using the movable incore detectors to obtain a power distribution sep: , c
- a. Prior to operation above 75E of RATED THERMAL POWER after each fuel i
, loading, and *
- 6. At least once per 31 Effective Full Power Days. . .
- c. The sensured F" g shall be increased by 45 for measurement uncertainty.
B i
. . . ~4 .
l. i O 4 e b 4 em
, ,, ,,.. --+,-2.m-eew----wr---- '*** " * *"" '""~ ~ ~ ~ #
_ . _ _ _ _ _ . _ _ _ _ . _ __ . . _ - ~ _ . . _ _ _ _ _ . . - . _ _ . . _ _ - _ . _ . _ . __- _. 6
, POWER DISTRIBUTION LIMITS LLANCE REQUIREMENTS dnved) _
da.
..Pr to ration above. TED' THERMAL fter *.T-T ' ' -
e
. . each.. f.
.. . end.ing, a.n.d
- 6. At least once 1 Effective Full Power -.. . . . . .
4.2.3.3 The sted RC5 total rate shall be verifi the region of accep le be within ' ration of to 3.2-3 at least onc e 12 hours ~ values of R g- , obtained per ification 4.2.3.2, a
*~ '
ssumed. 4.2.3.4 RCS total flow rate 1 itors shall be s d to a CHAPNEL CALIB at least once per ' nths.
. 5 The RCS flow shall be detemined asurement at least per 18 months.
b ( 4 A SEQUOYAH - UNIT 1 3/4 2-12
, . G.Om W.
d .M r
. - . ,.. .. - - .- ..- ... . .-. .. - - - - - - . - .-. _. - .- -._. -. - - . - ~
u ... 1 I' L m , 9W e . N . (x 1 L m_ _. ._ , i
--.. a =====: ns , ,,,,_ , .
j gggigglllgg I ; _,,7~; ;; 5 !E!EiEEE!E rB ;=
- I s saann;ssss sEEE'sssse l
.t: i EEEE sass- 5 u a : -
?$
2 k N 5l MIF - e*, g -
====== ===
= .
s --= ---=. = = m a;551 O J-i
.v . .
.=--_< E l -. --..--
3 g __ -w S, = ; l[ _gg U *9 t q) = ___.s -)- I ,
- _p <
- ____=
L (d '=~== " - = = = = =
=fiEiisis %. S
! ,< - sassessesssssssses N i , .- mi { $g, , O a.f.f. t i= E L s. g 8 V . . W-E
- t. r .sg.
. ==;=== s_ 2 ,
_ aM s
--- 88 ja F c, O { 8* *,, .
l5 g a F
- 4 g L' +_ :
'l a
. t .F.. -.
__. - 9 q lIy I, I-." - . = = . _ , . _ _ _ _.__--
-t -
89 es g
; s an. 4 P-- - + ,i,,,,,
.s _..i. .q , ,
. e ;# .
- 2 4
/t1 e s 's a n
, }
8MDe04 RAVtMA01d17101CDs
- q. ' . .
December.23, 1982 . SEQUOYAH - UNIT I 3/4 2-13 f
. Amendment No.19
~
8 f - 4
'o .. . . .. .
4
-,_ . - - . - - - . - ,., --- ..----.-....---.,,.-.-.4--nu,,s .,a.,aax..,w.. n .,y..n.ua -a , ii.
g . _. : .. 5 .! g
-:C' Q ~
. . . . . 8. .
= -
m: . >
** u !.- ....
*!j *l' 9 ' I . ... ,
- r *: 1:t 9
m -- . Q tv . .
-m .:=
"I . i . . ,. ..
g.
... .:s . - -
, I s' 3.$ U '"*
1
, e T.II N ..I. il Ei' ." . i. . .h a
/t a a a a *
. SW 6
i
. m l:. .
SEN10YAH . UNIT 1 3/4 2-14 L
.+.,.m..- - , . e n., g,,, . - , , - . . . . -
. . . . - , . - , . . . ~ , - - , - , . . - - - - - -
POWER DISTR?BUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION
~-
iL :llid2KL .
' ^ ~ ~
3.2.5 The following DNS related parameters shall be maintained within the limits shown on Table 3.2-1: - a. React w Coolant System T,yg.
- b. Pressuriter Pressure 4 d , - k # cTOL dc;ocA Mi~ 5'/SIM i
. . AP_PLICABILITY: MODE 1 - - - - - - 8
- - - -..-.. .TbThL.Fu)W 0- WT6 ACTION:
'With any of the above parameters exceeding its limit, restore the parameter ts
., within its limit within 2 hours or reduce THERMAL POWER to less than 5% of
( RATED THERMAL POWER within the next 4 hours.
~
SURVEILLANCE REQUIREMENTS 4.t.S.I 4re d Each of the parameters of Table 3.2-1 shall be verified to be t e in their limits at least once per- 12 hours. 4.24.1 TH C k d 5 FW & ATE J N Au 66 b 6 TEMEMED E 7 ma asu,w:w ur er censt ~ o n e P s /t 18 moes. 4.2.5,5 [gg Ub TbTN- Ft cco (LM~G & MbECAM5 4 t+W-BE .suisTecret to ,4 dH A N AJEC d ALIBR ATroN Ar tv r once pe,t 18 m aras,
),
SEQUOYAH - UNIT 1 3/4 2-18 y .
p -
;;; n=~; ; ;
% ,y y_,
;.3 P - ;T-wp a 1-
- m. ., TABLE 3.2-1
,h
'E g
= DW PARAETERS I 3
i
!..x g t-e LINiTS
Q .'
+
c - . 3 . i,J e x
~
* . 4 Loops In : .i;l.- ,'- kes/ .
PARAMETER l' operation v .;.' "
,. Il Reactor Coolant Sys,tes T < 583*F
, , an I. -
- glI i Pressurizer Pressure ,
_ I' 3. 2220'psla* - p }1 _ 34 i j Reador Coolant System ime- p a 3Y84oo gem # !j;I j
.,.. ,[.
w . A " per Limit notorapplicable minute duringstep a THERMAL POWER either in excess a THEM4Lof 10% RATED POWER ramp THERMAL POWER, in excess of 55 physles test, orlper RATED THEfMM. POW u of surveillance requirement 4.1.1.3.b. ~ ~ t. d ~ $Iil
# Includes a 3 5 % flow meosardment uncertainfy .
<j[
~
t* s n q
' ;j
*t i
e
- .M .
l En - . zu .
,o - e .- 9
~ . . .
Yy I- . o, - , ..
1E t u , :. {+ l K
- sl4.z.2. unci 314.2.5 HEhT FLUX hMD flUCLEhR ENTHk '
Q WOT CHA)HJEL-/FQ(a) adFS .
- l. - )
-h .
'3/4.2 POWER DISTRIBUTION LIMf75 *
. l n
3AsEs !!' /* 1 i i 4
~'
,, 1 The specifications of this section provide assurance of fuel integrity duringby:Condition events I (Normal Operation) and !! (Incidents of Nederste F -
-during a) asintaining noma (l operationthe calculated andDNBR in short in thetem core transients, at er above desig
- an fission gas release, fuel pellet tem t to within assumed design criteria. perature and cladding mechanical properties !
l In addition, lietting the peak linear . power density during Condition I events provides assurance that the initial . conditions assumed for the LOCA analyses att est and the ECC5 acceptance3 criteria. limit of.2200*F is not exceeded. 3 h these specifications are as follows:The definitions of certain het cha FS(Z) , Heat Flux Not Channal Factor,'is defined as the maximum local heat flux on the surface of a fuel rod at core elevation 2 div
-tolerances on fuel pellets, and rods.by the average fuel rod .
l F$g imegral of linear power along the red with the average red power. the highest intNu egrated power to m :' . . , 1 3/4.2.1 AXIAL FLUX DIFFERENCE (AFO) The lia'its on AXIAL FLUX DIFFERENCE r q assu'e that the F (2) up anvelope of 2.237 times 'khe notea11aed axial peaking factor is not excee during either normal operation or in the event of aenon redistribution follow
'ing power changes.
the plant process computer.through the AFD Monitor Ala The computer deter-eines the one minute average of each of the OPER *
- of 3 OPERA 8LE exws channels are outside the allowed AI-Power opera end.the THERMAL PuWER*is greater . .. -
than 50 percent of RATED THERMAL P
->- Y 4.2.2 and h .2.3 NUCLEAR ENTHALP% HEAT RISE NOT FhCHANNEL X HDT\FACTOR CHANN ' k' FACT 0e RC b LOW .
~
\ ,
i enthalpy rise hot channel factor ensure that 1). tw
- f. local power density and minimum DNSR are not exceeded and 2) in 't
' t
; of acceptancea LOCAcriteria the peak limit. fuel clad temperature will not exceed the 2200'F EC '
December 23, 1982
**%* SEQUOYAH'- UNIT 1 i 8 3/4 2-1
.4mendeent No.19 ee
.4;. , r
f q l L. ,- - o ! l POWER DISTRIBUT10N LIMITS g hggg , BASES '
~
1 Each of these s measurable but will'nomally only be determined y' ' ' : periodically as specified in Specifications 4.2.2 and 4.2.3. . This periodic i surveillance is sufficient to insure that the. limits are maintained provided m
- a. Control-rods in a single group move together with no individual rod i insertion differing by more than 113 steps from the group demand ;
position,
- b. Control. rod groups are sequenced with overlapping groups as described
.in Specification 3.1.3.6
. c. . The control rod insertion limits of Specifications 3.1.3.5 and 3.1'.3.6 are maintained.
\ /' d. The axial power distribution, expressed in terms of AXIAL FLUX Y \ DIFFERENCE, is maintained within the limits. R23 Fh w be maintained w hin its limits p vided conditio a through
. above are m ntained. As not on Figures 3.2- and 3.2-4, RCS ow. '
an 'F g may be "t ded off" against ne another to en re that the cal lated ()
/
11 not be bel DNBR the design D'N8R lue. The relax ion of Fh as a (. fametion f THERMAL POW allows changes the radial powe hape for all j pwaissible od insertion its. -
.\ N 7 a measured, no a tional allowance are When RCS flo rateandJAH neces ry prior to e pariso'n with' limits of Figur 3.2-3 and 3.2-4.
Measurem t errors of 3. percent for total flow rate d 4 percent for we been a owed for in ermination of e design DNBR va e.
, as calcu' ted in Speci.f ation 3.2.3 an used in Figure 3. -3, accounts for Fh ess than or qual to 1 49. This value is e value used in t various sa ty analys u ere Fg inf1 nces parameters ther than DNBR, .
peak clad te rature, and hus is the ma mum "as measure " value allowed. '
, as defined, llows for th inclusion of penalty for Ro ow on DNBR oni Thus, knowin the "'as mea red" values o F and RCS flo allow for H
" trade off" in excess f R equal to .0 for the pu ose of offsetti the Rod Bow DNBR enalty.
N ,. December 23, 1982 SEQUOYAH - UNIT 1 B 3/4 2-2
. Amendment No.19 9 E i
_ _ . . _ _ _ . . . . _ . . ~ - .- - ..
- i i
.:'a * ~,: * . - .: := .::..
= =:- _ . ;;. .. ~. s ..,___. r _
INSERT 4 u: :. . ..u.' A?$[;. . . . , . . . . . ve r ,r.:. . 'i The N (H as a function of THERMAL POWER allows changes in the radial power shape for a11' permissible rod insertion limits. - H will be maintained within its limits provided condition (sa-thrudabove,aremaintained. When;an Fg measurement is taken, both experimental error and manufacturing tolerance must be allowed for. The 54 is the appropriate allowance for a full core map taken with the incore
' detector flux mapping system and 3% is the appropriate allowanca for manufacturing tolerance.
When'(H is measured, experimental error must be allowed for ,
- and 44 is the appropriate allowance for a full core map taken with the incore detection system.
i. Thespecifiedlimitfor(Halsocontains an 8% allowance for uncertainties which mean that normal operation L , willresultin'FfH $ 1.55/1.08.1 The 8% allowance is based on the
- following considerations. '4
- a. abnormal perturbations in the radial power shape, such as L
fromrodmisalignment,effect(Hmoredirectlythan e F.o l -
- l. b.
although rod movement has a direct influence upon limiting Fg to within its limit, such control is not readily available to liinit F[H, and
- c. . errors in prediction for control power shape detected during startup physics test can be compensated for in F g by restricting axial flux distribution. This compensation for F[Hislessreadilyavailable.
e
~ =
lp,x
.L y
(
- F R D!sTRIBUTION LIMITS -
\,!
SAsts - M '
.. penalties applied to y account'for Red Sow (Figu 2-4) as n' function o are consistent with ,
(NRC) letter to. . Anderson (Westinghouse) described in Mr. John F.' 's April 5, 1979 and W 869
- I
, Rev.1 (partial' red bow ta)..
When measurement is taken, be risental error and manufe n tolerance must be a d for. 6 percent is the , riste allowenee for a .( core map taken with the ^ re detector flux mapping and 3 percent.
.is the . riate allowance for manu ~ ring tolerance. *
{ ll. l R23 i p .
.The hot channel factor Fq (z) is measured periodically and increased by a
(\')Iq . cycle and height dependent p'ower factot*, W(I), to provide assurance that the
,{. limit on the hot channel factor, Fq (z), is set. W(z) accounts for the effects of normal operation transients and was determined from expected power control maneuvers;over the fu11 range of burnup conditions in the core.. The W(z)
)'
function for normal operation is provided in the Peaking Factor Limit Report' per.$pecification 6.9.1.14. ~ i! 1 . u 3/4.2.4 OUADRANT POWER TILT RATIO
'l The quadrant power tilt ratio limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during startup testing and periodically during power operation. i 1 Thet)wohourtimeallowanceforoperationwithatiltc6nditiongreater s than 1.02 but less than 1.09 is provided to allow identification and cor- '
. tection'of a dropped or misaligned rad. In the event such action does not
- correct the tilt, the margin for uncertainty on is reinstated by reducing the power by 3 percent from RATED THERMAL' POWER r ea:h percent of tilt in
- j. excess of 1.0. -
o ( . / . .
, 1
)
December 23, 1982 , I SEQUOYAH - UNIT 1 - 3 3/.4 2-4 Amendment No. 19. Q . 6 9 9 l l
y, ; n, 1*. n 1.~. : ".:= ..a_._.,. ~ ~
= .; -:-
~- -
.7. ~_ . :
.. - . - --.-ZMSERT 5 -- - -- - - --- ---
w.~:.i . . n i. w.~u.- .m :c r. rw
. .-.. .~ ....-.
- p. .
Fuel rod bowing reduces the value of DNB ratio. Margin has been . retained between=the DNBR value used in the safety analysis (1.36) and the WRB-1 correlation limit.(1.17)-to completely offset the rod bow penalty, rc - The applicable value of rod bow penalty is referenced-in the FSAR. Margin in excess of the rod bow penalty is available-for plant design
' flexibility .
4 f
')
( ( .. U WD t S-4 4 S 9 4
. , . . , - . -- =
.g.. ,__
. - - - - - - - - - - - - - - ^ ~ ~
gg o-
- ;. . 7 q
t .
.o 4 - . .s 93 , :, ,-
p K
!,il -j (g
' POWER O!$7R!buf!ON LIMITS. *- j i
sAsrs -~ g e 4 M
, 3/4.2. 5 DNS PARAMETER $'
(
"' y '
meters are maintained within the nomal staasty state e assumed with the initial in the FSAR transient assumptions and accident and analyses. The limits are consistent' . . 1 j
-adequate ta maintain a minimum DNBR been analytically demonstrated-roughout each analyzed transient. '-
R23 Toadout is sufficient to ensure that the parameters . er are 1 limits following lead changes and other expe-ted transient operation. greater than-or equa.lsto the i safety _ analysis DNBR limit i t- s
. . t 1 ,
-J i
.. . 1 I .
. - ['-
- SEQUOYAH - UNIT 1 December '23, 1982
- i
. S 3/4 2-5 Amendment No. 19 6
' ? .E . '
..mf. ..: du s.p .... a. we.ened.. r -- - - - - - -
- _i _
---..gg,%. m . .% oe. 1
_. L<! , 7,p ..-w- -
I s 3 ;.. , ; 3/4.4 O REACTOR COOLANT SYSTEM-H k 1 , BASES HG .skFETy hMALy.5 % h Y t.swLT
, ,e , .
e' l ? Lf 3/E4.1 ~ REACTOR COOLANT LOOPS AND COOLANT CIRCULATION
. The plant is de's'igned t ite with all reactor coolant loops in opera-4:1 a, : tion, an'd 'aintain m DNBR'abov
,duting all normal operations and anticipated transients. In MODES 1 and 2 th one reector coolant ' loop not in operation this specification requires that the plant be-in at least + r STANDBY within
'I hour.
~In MODE 3, two: reactor coolant loops-provide; sufficient heat removal
?
. capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity.if a bank withdrawal accident can be prevented, i.e., by R88 opening the Reactor Trip System brukers. Single failure considerations require that two loops be OPERALLE at all times. :
?
In MODE'4' a single reactor coolant loop or residual heat removal (RHR) loop provides sufficient heat removal capability for removing decay heat; but: R16 single failure considerations require that at least two loops be OPERABLE. Thus if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE. ( In MODE 5,-single failure considerations require that two RHR loops be OPERABLE. n , EThe operation of one' Reactor Coolant Pump or one RHR pump provides adequate flow to. ensure mixing, prevent stratification and produce gradual reactivity i changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be
. within the capability of operitor recognition and control, t
3/4.4.2 and 3/4.4.3 SAFETY AND RELIEF VALVES The pressurizer code sa'fety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed , i to relieve 420,000 lbs per hour of saturated steam at the valve set point. l The relief capacity of a single safety valve is adequate to relieve any over-pressure' condition which could occur during shutdown. In the event that ric SEQUOYAH - UNIT 1 B 3/4 4'1 Amendment No. Itx 84 September 22. 1988 I
j' ' . . ' g;s INDEX-7 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIR !
',, SECTION..
=. . . L .
_PAGE ! 3/4.2- POWER DISTRIBUTION LIMI.T.S. -- 13/4.2.1 AXIAL FLUX DIFFERENCE...................................... 3/4 2-1. 13/4.2.2 HEAT FLUX HOT CHANNEL FACT 0R.............................. 3/4 2-4 , 3/4'2.3. 000 T' ~. l.T: f.O NUCLEAR ENTHALPY419E HOT bg CHANNELFACTOR...faw..................................... 3/4 2-8
.3/4.2.4 'I
.. QUADRANTPOWERTILT-RAT 10................................. 3/4 2-13 '
3/4.2.5 'DNB PARAMETERS.........'................................... 3/4 2-16 3/4.3- INSTRUMENTATION l 3/4.3.1 l REACTOR TRIP SYSTEM INSTRUMENTATION. . . . .3/4 . . .3-1 n 3/4.3.2' . L. ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMEt,TATION......................................... 3/4 3-14 E i 3/4.3.3 i MONITORING INSTRUMENTATION . i i:
-Radiation Mon'itoring' Instrumentation...................... 3/4 3-40 Movable Incore L9tectors..................................
.1 3/4 3 i Sei smic. Instrumentation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Meteorological' Instrumentation............................ 3/4 3-48 .
-1 Remote Shutdown Instrumentation........................... 3/4 3-51 1 Chlorine Detection Systems (Deleted) 3/4 3-55 R54 !
Accident Monitoring Instrumentation....................... 3/4 3 1 Fire Detection. Instrumentation............................ 3/4 3 Radioactive Liquid Effluent Monitoring Instrumentation.... 3/4 3-68 Radioactive Gaseous Effluent Monitoring Instrumentation... 3/4 3-76 N. SEQD0YAH - UNIT 2 V Amendment No.'54 October 30, 1987 t
p, ,
> ' ' .3';. Ef.'. ,
t i w-
- y
?w
- 3 E-INDEX.; l y '
w; . b ' sASES
. i
' . . . . . _ . . . . . _ . i
,f' ' S ECTION --
.,+r J4 PAGE l~ 3/4.'0 APPLICABILITY................................................ g' B 3/4.0-1 i i
, , 3/4.1' REACTIVITY CONTROL SYSTEMS I
- c. e *
[k ~ 3/4.1.1! B0 RATION CONTR0L.............. ........................... B 3/4-1-1
; 3/4.1.2 .BORATION SYSTEMS...... .................................... B 3/4 1-2 v, t h-
- 3/4.1.3 MOVABLE CONTR0L ASSEMBLIES................................. B 3/4 1-3 x.y '
JP - 3/4$2'POWERDISTRIBUTIONLIMITS u l.g w f
' 3/4.2.1 TAXIAL FLUX-DIFFERENCE....................................... B 3/4 2-1 i d
. - 3/4.2.2 and 3/4.2.3 HEAT FLUX ::07 :: 2 . T;,; G R . TJ : T LO'^t,T '
1 *
.AND NUCLEAR L 4
POWER_ENTHALPY TI LT - RATI0. . . .4. .M . . .BE.3/4-2-5.HOT
. . . . . .CHA
, l,
- 3/4.2.4 QUADRANTJ .....
C. 3/4.'2.'5^ DNB PARAMETERS........................ B 3/4 2-5 4j l
' J
, 3/4.3' - INSTRUMENTATION: ,
N/4;3.1 .and'3/4'3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY-FEATURE ACTUATICN SYSTEM INSTRUMENTATION................... B 3/4 3-1 .
!3/4.3.2 ENGINEERED S'AFETY FEATURE INSTRUMENTATION................... B 3/4 3-1. .
3/4.3.3 MONITORING INSTRUMENTATION................................. B 3/4 3-2 - 13/4:4 -REACTOR COOLANT SYSTEM ' y
- 3/4 4.1 REACIOR COOLANT LOOPS AND-COOLANT CIRCULATION.............. B 3/4 4-1 n
3/4.4.2~ and~3/4.4.3 SAFETY AND RELIEF VALVES....................... B 3/4 4-2 ' 3/4.4.4- PRESSURIZER................................................ B 3/4 4-2 .
. .o, s -
- t 3/4.4.5' STEAM GENERATORS....;......................................
. B 3/4 4-3 4
.y ' (!
u i4+ ,
's
/I
' q !-
1 ' i i s
, , - -hp6
- - - --- - ----- - - n - - - - - - - - - - -- '-
y Q, .t k[3h ,, s ,
~, ,
3 ,
' S> '
F '( 2.1c 5AFETY LIMIT 5 g, Q -. . - , - - - . [o , s WRB l corTsia%n and W .5 c4rrdatAn b L t.1 REACTOR CORE cdN nswti:de the coat et WRB-1 ,
+- ,
- The restrictions of this safety limit prew'ent everheating of the fuel and'
,' possible cladding perforation which sould result in the release of fission- k t K' products to the reactor coolant. - Overheating of the fuel cladding is prevented
) +
- ty restricting fuel operation to within the nucleate boiling regime where the y-L -
D heat transfer coefficient is large and the cladding surface tamperature is S". slightly above the coolant saturation tamperature.
'^ -.[
' :r- .
Speration above the oper boundary of the nucleate boiling regime could
' result in excessive claddin i '
free nucleata boiling (DNS)g temperatures and the resultantbecause of the onset
~ sharp reduction of departure in heat transfer y coefficient. ONB is not a directly measurable parameter during operation and Wrefore THERMA. POWER a ctor Coolant Temper.ature and Pressure have been T . re sted ta.DNs througn tna- terree* W h developed-to predict the 0 DNS tcomrte64ea hasFbeen-lux and the li>catio@n of DNs for axially uniform
- E. -.and non uniform heat flux distributions.. h local DN8 heat' flux ratio, DNBR, p L defined as the ratio of the heat flux that would cause DNS at a particular core location to the local heat flux, is indicative of the margin to DNB. '
i ~ The minimum va of the DNBR durin (f hj,-g t 1 oper nel transients, eady state ope 4atjon, normal d anticipated tran nts is-limiteo- 1.30 This value co sponds to a-95 cent probability a 95 percent con
- enee level l that DNB wi ot occur and is osan as an appropt margin to DN operating condi r all
- s. -
b The curves of Figure 2.1-1 show the loci of points of THERMAL PWER ter Coolant Systes pressurc and average temperature for which the minimum, DNERReac 110
.is no less than 1 or the averare enthalpy at the vessel exit is' eaual to the enthalpy of aturated liquid.
" th sdety u@is 'IMBR M and a reference cosine with a peak.of 1.55 for a axialof 1.74power sha
. ee n a.
is include'! for an increase in Fg at reduced power based on the expression:
.F =1.55[1+0.3(1-P)]
bl < l where P.is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control , rod f insertion assuming the axial power imbalance is within the limits of the ib(delta 1) function of the Overtemperature Delta T trip. When the axial power l BR alance is not within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection i consistent with core safety limits. SEQUDYAH - UNIT 2 8 2-1 Amendment No. 21,104 May 5, 1989 a , , . . ~. . ;-
]
' 2. 2 0 IMITING L SAFETY SYSTEM SETTINGS s
f (T ' BASES'
-. .c (Manual Reactor Trio '
' T'he'Mensa'1' Reactor' Trip ifa redundant channel to the automatic protective R instrumer.tation channels and~provides as,nual reactor trip capability.
i Power Ranoe.' Neutron Flux - t ~
"The Power Range,$ Neutron Flux channel high setpoint provides reactor core 4
- protection against reactivity excursions which are too rapid to be protected
.' .by temperature and pressure protective circuitry. The low-set point provides redundant dow power. protection in the power range for.a power excursion beginning from 7 J' 3 The trip associated with the-low setpoint may be manually bypassed , when P-10 is' active (two of the four power range channels indicate a power
* ' level of abo d approximately 10 percent of RATED THERMAL POWER) and is auto- J4 6
'natically reinstated when-P-10 becomes inactive (three of the four channels -
indicate a power 1evel below approximately 9 percent of RATED THERMAL POWER).
. . . ~ . . . .. .._ . . ._ . . . .
[' -Power Ranoe Neutron: Flux. Hioh Rates , h .T The ' Power Range Positive Rate ~ trip provides protection against rapid flux L increases .which are characteristic of- rod ejection events from any power ; levet. Specifically, this trip complements the Power Rangt Neutron Flux 4 High and Low tri L u partial power.. ps to ensure that the cri g 's are met for red ejection from the safety analysis DN8R limit r
.. :The Power Range Negative Ra ip provides-protection to ensure that the L
minimum DNBR.is maintained above for control rod drop accidents. : At high .
- . power aisingle or multiple ~ rod drop accident' could cause local flux peaking which, when .in conjunction with nuclear: power being mairtained equivalent' L ' . to turbine power by action of the automatic rod controi system, could cause
[L an unconservative local DNBR to exist. The Power Range Negative Rate trip L will: prevent'this from occurring by tripping the reactor for all single or a multiple' dropped rods. g :Intermed{nte and Source Rance. Nuclear Flux B The Intermediate and Source Range',' Nuclear Flux trips provide reactor o core protection during reactor startu). These trips provide redundant protec-
-tion to-the low setpoint trip of the Power Range L
-SourceRangeChannelswill.initiateareactortrIpatabout10 Neutron Flux per counts 4hannels. The
- second unless manually blocked when P-6 becomes active. The Intermediate m SEQUDYAH - UNIT 2 B 2-3 Revised 08/18/8'7
~
?
I '
"" ='s::z::::- ;- --~ :. . . _ . . ~ . . ~ -
L~ L* *
. , - .- - -- ~
_g
... . - ~ ~ . .
;g, ,,
b o , l +- The DNB de... .... sign basis is as follows: i there must be at least a 95 t > percent probability that the minimum DNBR of the limiting roQ during Cordition I and II events is greater than or equal to the DMBR limit l of the this DNB correlation being used (the WRB-1 or W-3 correlation in application). 4 The correlation data DNBR limit is established
" the entire applicable-experimental set such that there isbased a 95 on percent probability with 95 percent confidence that DNB will not' occur-when they;:.
e miniana DNBR is at the DNBR limit, ., I I e 2 _ _ . . -.
.u
.. 1 . . . . . . . . .
G a 40 4 9 6 L v
+, 3 ,
, a 'fy 6
LIMITING $AFETY $YSTEM $ETTINGS s
-eses Pressurizer Pressure
~ ~ '
Eihe PresNizer High and. Low Pr' essure trips are provided to limit the pressure range 31n which reactor operation is pemitted. The High Pressure itrip is backed up by the pressurizer code safety valves for RCS overpressure a protection, and is therefore set lower than the set pressure for these valves n (2485 psig). - The Low Pressure trip provides protection by tripping the mactor 4n the event of a loss of reactor coolant pressure. 1 Pressurizer Water Level-L
.The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit ~was taken for operation of this trip in the-accident analyses; however its functional capability at the specified-trip.settingisirequiredbythisspecificationtoenhancetheoverall
.: reliability of the Reactor Protection System.
' Loss of Flow-7
- The Loss of Flow trips provide core protection to event of a loss of one or more reactor coolant pumps.. prevent DNB in the Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will
.. occur if the flow in any twii loops drop below 89% of nominal' full loop flow.
- Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 89% of nominal full loop flow. F
- latter trip will prevent the minimum value of the DN8k from going below 1 F during normal operational transients and anticipated. transients when 3 loops E lare in operation and the Overtemperature delta T trip set point is adjusted to
- the value specified for all loops in operation. With the Overtemperature
- delta the P-8 T trip set76%
trip at point adjusted to the value specified for 3 loop operation. R DNBR from going below D THERMAL POWER will prevent the minimum value of the. r y transients with 3 loop 'n 5)during normal operational transients and anticipated operation. the safety analysis DNBR limit 4 <
, SEQUOYAH - UNIT"2 8 2-5'
7
- t 1,A g ,
' , Q -
' 't ?
= REACTIVITY CONTROL SYSTEMS 1 4 I
, w 4; R00 DROP TIME 1 2.7
,- y m LIMITING CONDITION FOR OPERATION L
o . ..
,M s 3.1.3.4' The individual Tull length (shutdown and control) ), op time from C the. fully withdrawn position # shall be less then or equal to ggs seconds from 1beginning of decay of stationary gripper coil voltage to da.shpot entry with:: i a.
T,yg greater than or equal to 541*F, and b.: : All. reacto'r coolant pumps operating. APPgCABILITY: P Modes 1 and 2. ACTION: w a. c With the drop time of any full length rod determined to exceed the !' above limit, restore the rod. drop time to within the above limit prior to proceeding to MODE.1 or 2. - 14. i With the rod drop times within' 11mits but determined with 3 reactor coolant pumps operating, operation may proceed provided THERMAL
- POWER.
POWER is restricted to less than or equal to 71% of RATED THERMAL ' -r o.. i _ SURVEILLANCE REQUIREMENTS 4.1.3.4 measurement The rod-drop tima of criticality: prior to reactor full . length rods shall be demonstrated through !
- a. ,
. For all rods following each removal of the reactor vessel head, ,
'b. t Z For specifica11'y affected individual rods following any maintenance on or modification to the control rod drive system which could
, affect the drop time of those spot.ific rods, and ,
- c. At least once per 18 months." -
l
*For cycle 1, this surveillance is to be completed before or by August 5,1983, whichever is earlier. the next cooldown R20 ,
l i
\
# Fully withdrawn shall be the condition where shutdown and control banks are at a position within the interval of >222 and 5,231 steps withdrawn, inclusive, R98 C
SEQUOYAH - UNIT 2 3/4 1-19 Amendment No. 20, 98
. March 28, 1989
. . . . = . . . _ _ _ _ _ _ . . _ _ _ _ _ _ . . _ _ . _ _ . _ _ . _ _
w.. .
-c :
. v 3.3-3/4.2.3 MUCLEAR ENT}lALPY HOT CHAalEL FACTOR- F -
m, ,. ., i
~
~
/
' eeWEn ersnisvTroN trMITs 3.2.3. Fj shall be Iltn@ed by b ;
- E y e. b nh rtheAT h no n Allah Wienship: '
.=
.LIMITINGCONDI*(fjlF0nOPERATION .- i
.3 M ,- .
- 3. 2. 3 The c ination of (cated Reac ow rate and Coolant Sys (RC$) total *
,R2 shall intained wi in the regio of allowable h
- n 'spe tion shown o igure 3.2-3 e 4 loop one tion:
~ ~-
,1 .
1.49 [1.0 + 0 9x (1.0 - PN g
.g, s
k x-a' b.- =
-L1 NR8P (Su)J
\K AH AI*Sb '000*E(l*0*fh Q -P = "NL NR i BATFD THEDMAt DOWFB 18 , .
, F"g= Measured lues of Fg obtained by sing the able.
""~
incere date ors to obta n a power d tribution . asasured - lues of F" g hall be use to calcula ' Rs ce Figure 3.
-3. include measurement neertainti
.of 3. for. flow a 4% for in
~ remeasurenetofFka
- e. RSP (su = Rod e burnup;, Penalty a a function f region ave age s shown in igure 3.2-4. where a rep
- is deft d as those semblies w h the same n Ioading te (reloads) r enrichmeht (first co .
APPLICABILITY: MDDE 1 AQI,0,N: . . ' o th the comb ation of RCS tal flow r and Rj , outside th regions cceptable o ration 'hown s Figure 3. : . Wit.hin ours: . -
)
- 1. Eithe store the l and Rj , to within instion of CS total f rate abnve limi or I InSerI 2 a. educe mE ER and red mR to ie mnseso RATEo mE l
the Power e Neutron ux - High tr setpoint less than or ual to 55% nA1E0'THE L POWE within the. t 4 hears. *
- 7 SEP 2 81983 ' '
SEQUDYAH - UNIT 2 3/42 ' Amendment No. 33. . 1' ;$6... .
@u , _ l i 1 l- _ . - - ..:..... . .. . _ . . . .
.~~,...e.-
INSERT 2
- r _. . . . . . . _ . . .. ._ ..
.u,.. , _
j, .. J *.*.~.4 24 '.O* * "I~ ~ ~
,]__,___-,,, *- - - = .
With FIH exceeding its limits a. Reduce THERMAL POWER to. lass than 50% of RATED THERMAL POWER _.
. within 2
- hours and. rehace the Power Range Neutron Flux-Hig'h Trip Satpoints to 1 55% of RATED THERMAL POWER within the next 4
, hours,
- b. . ~ Demonstrate thrti in-co're napping-that FgH N is within its limit
.within 24 hours after exceeding the limit or reduce THERMAL : POWER C' to less than 5% of RATED THERMAL POWER within the ,next 2 hours,
.and c.
Identify and correct the cause of the out of limit condition l: prior to increasing T*ERMAL POWER above the reduced limit
-i
**"
- a sepan OP M MON any
.. .E *.* ."# I*. b .ve proceed,providedthat(His_demonstratedthroughin-core mapping to'be within its limit at a nominal 50% of RATED THERMAL '
POWER prior to exceeding this THERMAL-POWER, at a nominal'75% of RATED THERMAL POWER prior to exceeding this THERMAL power and ,
.within 24 hours after attaining 95% of greater RATED THERMAL POWER.
C
- - - ~ ' ~ ~ ~ ~ ~ ^~
~1
,s 3; .
< y -
\
.t
\
. . j 4 '
\
POWER DISTRIBUTION LIMITS . d;= *- .- AE10N,,; -
, tinued)
- b.. Lttithi 4 hours e nitially bel outside above limit verify 1
, _through ore flux ing and '
tal flow te compariso that .- the combine on of Ry , and RCS.to flow rate restored l thin the abo limits, o reduce THE POWER to l s than 5% 'af ; RA THERMAL within next 2 hours. ' l c Identify nd correct cause of out-of-lim condition for
,N to-increas THERMAL above the duced,THE
~
POWE,R lia - uired by A DN items a. . and/or b'. ove; subseg t POWER OPE 10N may y eed provided t the e ination of R , R and g 2 indicat RCS total low rate are amonstrate 'through inco flux
. m SP.ri 3 mapping an Res total ow rate coup ison, to b within the re ion
, ol' acceptabt operation on Figure .2-3 prior o exceeding t,
. . Iowing Telf POWER lov s:
1. N inal'505'o RATED'THE L POWER. - t- 1 2. A i nel 75% of R ED THERMAL R and f p
- 3. . Withia 2 hours of at ning great than or equa to 95% of RATED THE L POWER.
. -y .
5URV LANCE REQUI ENTS - -
' ' s
. s x 331 T N ,
revisions o Specificat n 4.0.4 are ' t applicab .
- 4. 3.2 The e ination of icated RC total flow ra and Rg ,. shall be .
- i. dete' ned to be thin the reg n'of accep le operatio f Figure 1
-3:
P
- a. ior to ope tion above of RATED I ding, and ERMAL POWER ter each f 1 .
- b. iAt le t once per 1' Effective ,
11 Power D s. 1~ h e SEQUOYAH - UNIT E - 3/4 2-9 - l'
q
- 1
- e. .: .,- .
I
- m , -
nser7 3 L a 0*( , . . . . . . ., _. .
'r , ;
-!DRVEILLANCE REQUIREMENTS
- l
~
4.2.3.1 The provisions of Specification 4.0.4 are not applicable. "'' " .
' 4.2.3.2 F"gshall be determined to be within its'11mit by using the movable i incore detectors to obtain a power distribution esp: ,
- e. Prior to operation above,755 of RATED THElpN. POWER after eh fuel
,, .Ioading, and 1 L b. . At least once per 31 Effective Full Power Days.
L c. N The measured-Fg shall be increased by 4 for measurement !- uncertainty. l ? i
\
i y . , v ( .
")
e s G
. g e
i e * ' e e = essa e W u-+e .- -- - - w
i 4
/
POWER DISTRIBUTION LIMITS T SURVETLLANCE REQUIREMENTS (Continue () e , ) $. 2.3.3. The'indicatet 5 total flow of acceptable oper . ti'6f Figuree. shall'be v fied to be r.re -3 at least o ht Lthe. the ao per 12 ho *vhen .
- tion 4.2; recently .
'are assumedobtained to ex va, is of ( and 2 *D**I"'d P
t 1**' 3;4 The MC . . tal' flow rate indica CALI TION.at least s shall be s eted to a L ' e, per 18 months, *
', .3 T $ total' flow 7 shal.1 be deterni by measu enen t least 9
, 9 4 6 I
9 4 a G e e C . .. i
- e e e 4
l- ,
. -j '
e , e e e O 4 F .
'A e
e
- O O
O g 9 9
. 's SEQUOYAH - UNIT 2 '
{ , 3/42-10
~
q e a G 9 e g 8 eg 4
- wt_ - - - - - - - - - - - - - - . - - _ - - - - - m- -_ a - e ' m'
x:; ,,
. ,p::m
_ . E
~
, ) , ?. i Delete .
= .
m - S f A
- b a ,
~/ . m -
i vivwes ENT U.Cvw s w ivv w , il yI ! l ! e ' 3.5 OR FLOW AND 415 FOR INCORE ' WIEA NT OF F,", ARE INCLUDEO i ! l E eines mE. t, ..-l g 4 l 9: '
':lllllllt lp 'l'zlllltttlllll':lll ACCEPTABEE -
y
- OPERATION .-
j REGtONFg.. .l
. 1; A, Ogd :
g '
. 3 ed j
e , gg . I UNACCEPTABLE . i b . . i
- f. OPERATION *,
O j- RE080N ' g -- j 4 a :l-I t
'W
. g 42 ; ACCEFTABt.E h q 3
g
- g OPF R ATION
* . .s REGnON FOR
.I, y R, & R / (1029, es.ssi
- O y '
4e 31- , i i , .. i y; r-
- m. --, e h , I l\
t' (1.0,37.94)
; UNACCEPTABLE i l I, O.ERATION ' t M ,
H REGION 4H
-[l LllMIHillHIBMlHR
.t 11!. -
g .se o.or o.se o.se o.se sm t.oa im 1.cs g-g a,-a,ni-. :
,,a FIGURE 3.2 3 RCS Total Flowrote Versus R, and R, - Four Loops in Operation am
. g .
N I
-am o " " ~
m I1 $,. y<Y . o, e. M t. ( . t
\
- w. --
=
)
I l _m- '***k e.
..=Mp.o-e.',
, .-w.- paap - ~
M.._, m
,q
.Oe *.
1 e
.e .m,,..M.. .e.&', N g -- --,
,.,,., ,_.e_
t .I *'*,,*~N
.'s.im..
en--w etsb e4 ae.mmum-
. ~ -
-, Of. _ .g1
_..e.e_ _ .
, **l=--
... g e. t
.+.ee.-
-r n.cT---
M .e .e.e .o-.
. e en.,
e..e.-, _ =2 aus g me. ee w .e . - _ - g.e Emm.
-e.e e
. e. M 9 iM .
w -**Y"--- ' m g
*m.
.e e.1""."e.e-.4 m .e
,.e e
.e .e.e.e. e .
e um. l
,~ . ,
..~e.-_* --$ '" - - -
unumus.emme. -- _
>GE..*
em e.ee eufgh 43. -'--- Mm.e ege e .muuWenn.e; e.e ee. m- Wep. N e e. . IE. mo .emmes asum. - ee emab=m ".m mau a..... ..e.....
. . . ..e.
em.,e...E.,
. . .es..-.
.e.
...eW.e.me. ...i+M~..e .
. .eu 9...
e
- 4 eeg .ee .m.. . .... _09 e.9..lui .b.eu_W e -
g
. . ...e...
e e. 4e.... es... .e e..e.... 9
...e-e ,
....e.
4.e..O e.ee...Oe.. e.Wm O. - - -_ e .eet. .m. 4 . W We m. eep W . 9 m.
-ee ...e 6e ,e.g. m a
e.y . ee..e... e
- e. m.peegg.h a y ee, 9 egg
- 6. Iwei.. . ,
am
. .. ..... .. e . ... .e.
.e.
..e9e eg. e es.e e
.. ..O.*499
.a w... .ag e .
g.g
.e e.
g a .gegg ..e...e.s.,. .
.. e ,ua
.e eye.e4 .e
- . r. ..e.e.r.
. - - - ge e .en.e o ggg . e e.e ee. ege.g
.e.e-...e
.e.
..e.,
. e. e 4 .e e ge,. , e .
4 6 6,. m% m.sgn.
*=.e.g.
.=e.
4 g . ye . *- Ie. . . g e ee . u-.* e gum.4.u. De ... 9 e e ...o= .e. aeee.S e e .eu.gebe. be.eee
. ee e g .ee se.e 6.e.emme y g. ge
- p. pge , e
. p.m e.g.-
geeg g 4 $5-we.gg ae+.eegre. see. e. e.esma e aus . e.m .e g . egg og. o g e .e emes e.gg es eg eme ee<e .e e4m, . gu. gh.
.. ...*9.Oe..
Se. .eeg Seu . . e . . eD . . e e . y.e. 9 e .t . ea
.. - -~ g ee 4e e e e . e e o. . . e .
ee.ee m. m o e Deeee e_e pe.e.e ee. eMOge.e em. he M e .e g .e ..e e. ep..eu
.g g e e.e. .eeee e..eisee. se es e* og e. e..
u
.eeu.g -
'd 5M I
WW e. e. em e. = em enum 4 re p seu.a. .
.M$
Geg . M e N .6e.M* i e- -- * . C
' a.e.em. e e e .a. e..96sse.' " see.ee e a e e _eh em @. 963 e . e e en... .e..he.ehWS.e.
. .e e Im.= e.. 6 e.
. h .Oe ....e. .m me-
- e. .
e...e. e g.
.e4.g...
e . eus..e.es.ema .. .t oe . .a. e.ne... eee em
, ee .e e an. ..ee
. . e.. -
o
.. ***mmenem
.e6. e ee...=d.
=
g . $. a. e. W home , " em. 9 6. 9..e. m W M. . .e..b 4..- 6..
. see. . ee. ,
ae p p. e .g e$ M
. '.e. * .
.M.Im @'-_ ue 99 ee -- e am. em M 9 $_ _.9 & . _ @
( amisms.un. .eimm.
....e".Te. . e. o
_.m. . me
.e.e.,
.e . eme.a a. . .- e.
..e.e.m. --
.e- . _ -
ee..-. e
- e. e-
.e.
e g
.e.
e-- e. _.e.,---U.* e g q tee 48 e ...- 4
. #.4 e
=e e. e. m. . . eeme9 - e em e e
- emme. em... *e* '
em.
== d *e. a em .
W
.eo=e....
- e. ... .o en e e' .
.eu.e samm.e s ' eemdu. . . 'eso. --- -
.SM"e._e_
eeee e -- - - - M N p g. w e
.eWee tee. e0 meeue. mS4 i ee .*e ea e --
sm e e' m..-- - ' - - M.@ M eq - 6_ 4
.M h eugMeM. ,e=
.h4 M -- M 6' 4.e.e en .e.e.,e...,e...T.r ...e. -----e u.m.m.uus e
. 7es em4me.u
.es..
gnee
, e o e e. e. e .e.e .e....e. e e
.e. ... e.. .
e.,,
- e. --
- eea
.ee on e +
e.D.e eg. go .i ee.
- e. e e een e ,. e. e e.p W ^
em e eo e.e .e .
.e..e.em.e.e e. e.es ee p. .w.m..e. e.ee. e.es. .ee..e=G.+.
g
...dg. . edummuis.e
. spenhe e ee meemshe e en emapu siee.e
. em s.e ge " semus .
- c 1.em em see. e mesa.mmum me.
. e.g.e.a.e .e eee. . . . eau. e . ejm eem eem . --
...m.e. m.*e ano.
e... --
** e. ***.e.eu. use.
b..e4.M...e. .e t. e. .e .BD N. 2 . e. g e. .e. ausee-
. . e.6.ee m. e. .
e.....
.e....e--
l
. .,.e. 0.
.e...
_ ..e._
.e.- - -
g me...
.... . .ee.,
.. in .e l
. .. . . . e.- ..e..o...e e.. e. ..
eg .a s ,i e,e....... gg. .e 9m..,. "' pp
.u W .
. eg.4.D.= 1.- #6 eW. a- _
l eeea.e. e, ., 1 m.,,
-4 h e 4 e q eg1 me se m o egu,,. "g ee.e m.g
..e.
pe..eg e . . . es p e
. a. eumm. e.a. es.m. num.
W- h m.
.. ==mme. sumum
. . .eee , ee g . o 6e4 e 'e . .s e. y e.
ammi
..ge sem e e.a em eme ee ee . . e6
-s
=-C mte
. _...e . . . . ..
. .. . .. .e..... .- _.= - - s.
..-... . .. ..... -. _ . . .-_ ...._e. . . . . .._..
e
. ...... . ..e..
. . . ..- .e.
.e_. .m. O. - . <
e.... -- .-.
.e..-.
I I
...e_
9. 4 4- E 4 E
- deu N
)1i
(. )i , SEQUOYAH UNIT 2 3/4 2-12 . I
- 1 1
\
,. a z, , . .
.j \'q t -
q ,
~~ y " . S'fi s'
+ -
7 _ , _
, - , t, ePOWER DISTRIBUTION LIMITS 1 h1 3/4'2.5 DNBRARAMETERST
, LIMITING CONDITION FOR OPERATION m
;F !3.2.5; The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1: .
a.. Reactor, Coolant System T,yg. . f . b.
. Pressurizer' Pressure.
WT' - TSN jAPPLICABILITY:. MODE 1. gg Q 9 A rf ' iACTION:- ' LWith'any of the above parameters exceeding its limit, restore the parameter to *
;within. its;1imit within 2-hours or reduce THERMAL POWER to less than 5% of
' RATED THERMAL POWER within the next 4 hours.
i
\
5 SUR'VEILLANCE REQUIREMENTS- - i d. ' I g 1 14.t 4,1 . . . 4r+r4.':Each of the' parameters of Table 3.2-1 shall be verified to be within their limits.at .least once per 12-hours. l$g1 {hg77pf2.vuSMLD
~
Ll 2..$.2. M(f - kb ' pg,oco ft,4.pg s p 4t t 3 f wcre s ad-E m e >tr A f t.gytsr o nes Pet 18 win s, b t 3 s : :.su.a.I screb rc, # d14AMet dAL r3 R ATrza.f av i
.LERST DRc 6 Peit i 8 M-oNTM.S ,
e 0 4'.
.SEQUOYAH - UNIT 2 3/.4 2-16 j i . .
Ii
'W k r
$# l lllll t IMAGE EVALUATION ///jp/ {od'#
4 )y g'A @ TEST TARGET (MT-3) yN'N[ j[,4@4#, 4
% > W' '%,,,, '
l l l l.O lW4 RM
'dllE4 '
1,i [mE?e l.8 l.25 I l _.4_ 1.6 l l 4 150mm >
< 6" >
ll . 4 % sp b 4 % t;p?.'6,,,,,
? 3,,,g gj<y;ja o>////,;
hs:E _ _ .--:----- 4"- U '> _- -- -)
e-
,3+N
# t$+ E <
pS $ g,77e%f.%I/ IMAGE EVALUATION TEST TARGET (MT-3)
/,,,p jd M
4 g>,+p %,,+y l 1.0 lfE2 E4 llj IL2g l l,l [m lLM l.8 1.25 1.4 1.6 4 150mm >
- 6" >
4;# % ,,,,, f5,,,,,~ . ,
/<q)f4 4%
% // i_ Ab**
g ;;_,,_ , _ . .
....i :
J
-3,-.
IMAGE EVALUATION d#4 4)%@ g f>d' ///// rest TAnoer (sim 7ppp g
+ .
l l
)
l l 1.0 lf 2 Ils 5 9 Ell E!Eg=2o I l.1 L l 1.8 1.25 1.4 1.6
?
4 150mm >
< 6" > #f:/ %,,,,, / 4%
Ar f7777 3<Qf4% bi s y . o~ . -. -
, s
3 E 5*
#d>O@' IMAGE EVALUATION ((j/ 8/4 F
4x restroest wrm pppp p I.0 E 2 EM 5!NE l 1 l'l E m RE l.8 l 1.25 1.4 1.6 4 150mm > 4 6" >
% +
- fh>,,,1,
~
& s
= -
m ,. mw +c;- mff[p3m:- r
-3, ,
.~; .;;; -
'r
,1 -
-':~' }l A 2' D . gl'
'. D
/ 7 ... p
, l+* =
^y 7.j.
p 2 t
. ...: ;_. q
=s_;. -
; 3..
m - ..
- v v sa E..
TABLE 3.2-1' c 4 - ~M i o = . f
- 5! .l 1 -: ; : . ::
x IDM8 PARAMETERS' n. ,l I i
~#
. 7 , , .
# ..i* .
~
. 2 it . .
4 ' -L M TS- .I
.s
. u ^
6 - - ' PARAMETER
, 4 Loops In - -
. operation .
. E3 . . i;.
/ -j ; .
~>
Reactor Coolant System T < 583'F.
~
avg - . . - o Presserfre,r Pressure- > 2220 psfa* -
~
4
- ~ - -
~
Etodor (oohnt SystemF fort a 3784oo3pm a O
~, -
D* . > M .
- ,!i .-
. . o ,, .
j,
~
f* .j G. g * , J);
...~-
)
. .J . -
.i
. E. a . .
- 2. _
==
*. l.,. -
g:u . uG
- Limit not appitcable during either a THERMAL POWER ramp in excess'of 5% of RATED THER M L ' '
W3 - ' POWER per minute or a THERML POWER step ~ In excess of 10% of RATED THERMAL POWER physics ' test, or performance of, servefilance requirement 4.1.1.3.6. ' l# Includes a 3 5% h meomremea+ nan AaM
.- 1 1 ~-
l 233 >
a. I t
~
).. * . FLUK MD NucLEkR ENTHALPY 5l4.z,z. ank sl4.t.5 phT A '
WOT CHANMEL -Fem) and F - ([L' - 3/4.2 POWER 9157RIBUTION LIMITS BASES --- ~ - "~ e . The'spect:ications of this section provide assurance of fuel integrity
, '.. events during by: Condition 1 (Normal Operation) and II (Incidents of M ,
during noma (l operation and in short tam transients, and ' to within assumed design criteria. fission gas release, fuel pellet tempe In addition limiting the peak linear power density during Condition I events provides assurance that the initial criteria limit of 2200*F is not exceeded. conditions assumed for the L ' theseThe definitions of specifications arecertain hot channel and peaking factors as used in as follows: FN (I) ' Heat Flux Hot channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided
'by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods. .
F" -
-18u Nuclear Enthalpy Rise Hot Channel Factor is defined as the ratio of the "
i' ((~ . the average rod power.tegral of linear power along the rod with the highest i
- 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)
The limits on AXIAL FLUX DIFFERENCE assure thatq the F (Z) upper bound anvelope of 2.237 times the normalized axial peaking factor is not exceeded during either permal operation or in the eve'nt of xenon redistribution follow-n ing power changes. , the plant process computer through the AFD Monitor Alarm. Pr The computer deter-
', mines the one minute average of each of the OPERABLE excore detec -
and provides an alare message tamediately if the AFD for at least 2 of 4 or 2
- and the THERMAL POWER is greater than 50 percent o AM.2.2andh4.2.3 HEA UX HOT C EL FACTO CS FLOW AND NUCCEAR ENTHALPERISE HOT CHANNEL FACTOR \ \
A. I The limits on heat flux hot ' channel factor. IRCOTdWP3t1D an enthalpy rise hot channel factor ensure that 1) the design limits on peak s ( "~ local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peakItait. fuel clad temperature will not exceed the 2200*F ECCS Ng acceptance criteria - SEQUOYAH - UNIT 2 8 3/4 2-1 Amendmenc.No. 21 i
- SEP 2 M383 i
L L . _ ._
e m
, l q ,
l l
.- Y _. , _ , ,
l R DISTRIBUTION' LIMITS j WE/ #90 #II BASES
. .- f']'
J Each ofit:hes/is measurableallybut will, no 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: a. Control rods in a single group move 'tdget6e7with no ' individual ' rod d insertion differing by more than + 13 steps from the group demand position.
. l
- b. - Control red ~ groups are sequenced.with overlapping groups as described
- in Specification 3.1.3.6.
r .
- c. The control rod insertion limits of Specifications _3.1.3.5 and 3 3.1.3.6 are maintained.
- d. The axial-power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained'within the limits.
Fhwill maintained within s limits 'provi conditions a. rough-d.- ove are main i ned.- As noted on ures' 3.2-3 and 2-4, RCS flow. H. and Fg y be " trade off" against'one an her to ensure t t the calculate "
~
DNBR will t be below t design DNBR value. The relaxation o Fg as a ' ction of T MAL POWER al ws ch'ande's' 'in 'th'e' 'r tal power shape or all perm ible rod i rtion limits.
^
When R flow rate a Fharemen reci,' no . additional lowances are cessary prio to compariso ith the lie of Figures 3.2-3 3.2-4. N l Meas ement errors f 3.5 percen for RCS tota low rate and 4 perc tforFh l- have be allowed for n determinat of the des DNBR value.
, R as c culated in Sp ification 3. . and used in igure 3.2-3, accoun L for N),less than r equal to 1. 9. This valu is the value ed in the various afety analy s where Fh luences param ers other tha DNBR, e.g.
p peak clad operature, d thus is the ximum "as mea red" value a wed.
. Ts^defliie ~ alloids fo'r' e' inclusion ~a penalty for d Bow on DNB on1 Thus, kno ng the "as sured" value fFhandRCS ow allow for W
t-
" trade ff" in exce of R equal 1.0 for the rpose of offset ng the Rod s
. Bow DNBR enalty.
l SEP 2 91983 SEQUOYAH - UNIT 2 8 3/4 2-2 Amendment No. 21 l l .__-- - - -
z . i 4 " Up es . W = $ 4-c
- . . . . . - . . . .....:. =-- , =- - -; ~., . . ~; : ~ ~ 1 INSERT 4 '
l
- g,t Nt r u -
The F[gasafunctionofTHERNALPOWERallows changes in-the radial power shape for all permissible rod insertion limits. - H will be maintained within its limits provided condition (sathrudabove,-aremaintained. l l 1 When an Fg measurement is taken, both experimental error and l
.. ~ manufacturing tolerance must be allowed for. The 5% is the appropriate allowance for a full core map taken with the incore .
detector flux mapping. system and 3% is the appropriate allowance for
; manufacturing tolerance.
When(H.ismeasured,experimentalerrormustbeallowedfor , and 44 is the appropriate allowance for a full core map taken with the incere detection system. Thespecifiedlimitfor(Halsocontains ' an 8% allowance for uncertainties which mean that normal operation willresultin(H $ 1.55/1.08. The 8% allowanc.e is based on the following considerations. v l l
- a. abnormal perturbations in the radial power shape, such as from rod misalignment, effectFfHmoredirectlythan F. g
- b. although rod movement has a direct influence upon limiting Fg to within ite limit, such control is not readily availabletolimitF[H,and
- c. errors in prediction for control power shape detected during startup physics test can be compensated for in F g by restricting axial flux distribution. This compensation for N
F H is less readily available.
o . r
,. l. ,
- ' + .--,
F/ , . POWER O!STRIBUTION LIMITS . .. ,
. BASES' 4 enalties applied to g ouniforRodBow(Figure .- as s' L function of are consistent with those ribed in Mr. John F. Sto
. (NRC) letter to T. . erson (Westingnouse) date il 5, 1979 and W 8691 -
, v.1-(partial rod, bow test . .
When an Fq rament is taken, both exp ntal error and manufactur n tolerance must be allowe 5 percent is the approp allowance for a ful map taken with the incore ctor flux mapping system t 3 percent , is the approp ailowanceformanufactur lerance. ,
.+
; t -
M
, In.Ser, The hot channei factor Fq (z) is measured periodically and increased by a cycle 'and height. dependent power factor, W(z), to provide assurance that the ;
{ limit on the hot channel factor, Fq (z), is set. W(z) accounts for the effects-of normal. operation transients and was determined from expected power control g ;; maneuvers over the full range of burnup conditions in the core. The W(2) ' function for normal operation is provided in the Peaking Factor Limit Report ) per Specification' 6.9.1.14. *
~3/4.2.4 0UADRANT' POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during startup testing and periodically during power operation. The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and cor-rection of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on Fnis reinstated by reducing the power by 3 percent from RATED THERMAL POWER f5r each percent of tilt in
. excess of 1.0.
SEP 2 91983 . SEQUOYAH - UNIT 2 8 3/4 2-4 Amendment No. 21
)
i i l
. . , i~ ~ T"7.'"L. . . O;L .-- -;; - - -
. J- --, *-
T - r. zus e s m .. -. . 1 i
- n. ... a v. : .
.. . .d u rc:
.. w.- . . : a:- , .. - . .' ant l Fuel rod bowing reduces the.value of DNB ratio. Margin has been retained between the DNBR value used in the safety analysis (1.38) and )
the WRB-1 correlation limit (1.17) to completely offset the rod bow i penalty. The applicable value of rod bow penalty is referenced in the FSAR.
' Margin in. excess of the rod bow penalty is available for plant design flexibility.
I t
~
1 I i 1 P e
f * ' D< *
, g, , . .
POWER DISTRIBUTION LIMITS- , ,
. t;!r BASES
.y 3/4.2.5 DNB' PARAMETERS- -
meters Thearelimits on the DNB related parameters assure that each of the para-
, : - ' I';-assumed in,, maintained within the normal staatty state envelope of' operation '
theFSAR with the initial transient assumptions and accidentand analyses. The limits are consistent adequate to maintain.a minimum DNBR been analytically demonstrated r ~ hvoughout each analyzed transient. ' g
. The 12 hour periodic surveillance of these parameters through instrument
' readout is sufficient to ensure that the parameters are restored within their '
<s limits following load changes and other expected transient operation.
. greater than or equal to the .
j safety analysis DN8R-limit - p * ( 1. i. t p I l l: . 9
- i. .
L1
- g. . P L SEP 2 91983 SEQUOYAH - UNIT 2 8 3/4 2-5 Amendment No. 21 4 w ..:. = % .. ... . a .a -- -
; wa -
_. ..w .., - . a .. a
4
,1 4
~
1 3/4.4 REACTOR COOLANT SYSTEM i _ , , , . , y 5Af6T7 Adh'Y4 U . N .b.- p r T gAus:
/ ..
3/4.4.1 - REACTOR C00LANTJjp?S AND COOLANT CIRCULATION The plan't' is des'igned to oper te with all reactor coolant loops in operation, and maintain DNBR 'abov anticipated transients. during all normal operations and In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the' plant be_in at least HOT STANDBY within 1 hour. In MODE 3, two reactor coolant loops provide sufficient heat removal
capability for removing core decay heat even in the event of a bank withdrawal-
' accident; however, a single reactor coolant loop provides sufficient heat R75 removal capacity opening.the Reactor if a bank Trip withdrawal System accident can be prevented, i.e., by breakers. i require that two loops be OPERABLE at all times. Single failure considerations-In MODE 4, a single reactor coolant loop or residual heat removal'(RHR) loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.
Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE. In MODE 5 single failure considerations require that two RHR loops be Q OPERABLE. L flow to ensure mixing, prevent stratification and produce gr changes during boron concentration reductions in the Reactor Coolant System. l The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. SEQUOYAH - UNIT 2 B 3/4 4-1 Amendment No. 75 September 22 1988
ni u
^%. ,
,t 7 ;[ 4 qll. '
.,l '-
,g c a
- 5 \. .. -
s if
;y v ,
.' ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328
, . . . .. ., ..c- . , .: l
, , . (TVA-SQN-TS-89-33)
DESCRIPTION AND JUSTIFICATION FOR VANTAGE 5 HYBRID FUEL UPGRADE l ip 1 1 4 9 e w iW P i l_ l
^
fi' s I i1 ENCLOSURE 2 !
' Description of Channe The: Tennessee Valley Authority (TVA) plans to refuel.and operate'Sequoyah Nuclear Plant (SQN)-Units liand 2 with Vantage 5 Hybrid (V5H) fuel that L ' incorporates low-pressure-drop:sirealoy grids and removable top nozzles.
[' integral' fuel' burnable. absorbers, and extended burnup capab111 ties. This-
' upgraded fue1~will also contain debris filter. bottom nozzles, snag resistant' grids, and standardized pellets. The evaluations performed for
.this fuel upgrade accommodate;the effects from the following modifications planned for the Cycle'4 outages for each unit:
M 1. Resistance temperature' detector bypass elimination
- 2. Eagle 21 digital protection system
- 3. Upper head injection removal
- 4. Boron injection tank removal
, 5. New steamline break protection ! 6. Reactor trip on steam flow / feed flow mismatch l As a. result of this fuel upgrade, TVA proposes to modify the SQN Unita 1 and 2 technical specifications (TSs) to revise the bases for safety limits to change the W-3 correlation to the WRB-1 correlation and to revise the associated departure from nucleate boiling ratio (DNBR) limital to revise TS 3.1.3.4 to incorporate a new rod drop time of less than or equal to - 2.7 seconds; to revise TS 3.2.3 to delete the rod bow penalty as a functionofburnup'intheF[H(NuclearEnthalpyHotChannelFactor)- . equation and delete Figure 3.2-3; and' revise Table 3.2-1 of TS 3.2.5 to define' departure from nucleate boiling '(DNB) related reactor coolant system.(RCS) total' flow' rate limit, including uncertainties, to be 378,400 gallons per minute (spm). Reason for Change The change of the W-3 correlation to the WRB-1 correlation and the revision to the design DNBR limits in the bases of the safety limits and the increase in the minimum rod drop time are required to allow implementation of the improved fuel design for V5H fuel. The deletion of rodbowpenaltyasafunctionofburnupintheF[Hequationand deletion of Figure 3.2-3 reflect new evaluation methodologies for the i effect of fuel rod bow on DNB. The new methodologies provide a basis to L eliminate unnecessary power distribution penalties and to simplify the
- specification. Relocation of the RCS flow rate requirement from TS 3.2.3 I
to TS 3.2.5 is also the result of new evaluation methodologies for the effect of rod bow on DNB. This change clearly defines the DNB flow parameter limit. l [ I l
y m i g- j l l LM 1 Jn summary the proposed changes are primarily the result of the following I three items: . i
- 1. Use of a new DNB: correlation
- 2. Increased rod drop time because of the reduced guide tube diameter for V5H sircaloy grids ,
- 3. Incorporation of current methodology to assess the rod bow penalty Justification for Channe
.As discussed in the safety evaluation for the fuel upgrade (Enclosure 4),
the previously reviewed and licensed safety limits for SQN are met with the upgraded fuel. The new fuel design has provided satisfactory operational performance.in fuel assembly-demonstration programs since the early 1980s. 'The V5H fuel is both mechanically and hydraulically compatible with'the current SQN fuel assemblies, control rods, and reactor internals interfaces. The V5H fuel satisfies.the current design bases for SQN, and it meets design requirements for hydraulic stability and structural integrity under. seismic / loss of coolant accident (LOCA) loads, with margins comparable to , 17 x 17 STD (stands.rd) fuel assemblies. Nuclear characteristics are s comparable within the range normally seen from cycle to cycle because of fuel management effects. . No change in fuel rod design criteria, methods, or model is necessary with ' transition to V5H, with the exception of a new DNB correlation. Based
- upon the information provided in the evaluation, the SQN plant operational limits will be satisfied with the proposed changes.
The evaluation considered the effects of the proposed TS changes on the following areas: l 1. Mechanical, nuclear, and thermal-hydraulic fuel assembly design
- 2. Non-LOCA accidents
'3. LOCA accidents
- 4. Environmental consequences of accidents These areas have been evaluated for the impact of all proposed changes, I
including the transition core effects (with a mixed core fuel loading with both V5H and 17 x 17 STD fuel). The required analyses, as described in the fuel upgrade evaluation, were performed by Westinghouse Electric Corporation using methods and procedures previously approved by NRC.
- l. DNB Correlation Change .
The calculational methods currently used for 17 x 17 STD fuel assemblies and described in the SQN Final Safety Analysis Report (FSAR) are applicable to V5H fuel assemblies, except for the DNB I- correlation. The new correlation basis for DNB performance is the WRS-1 correlation.
a l o I
.c The WRB-1' correlation established a DNB limit that provides for the
- margin of safety: required _by the current FSAR (i.e., DNB will not
<. occur on'at least 95 percent of the limiting fuel rods during normal li, and operational transients and any transient condition arising from faults of moderate frequency at a 95' percent confidence level). The WRB-1 correlation takes credit for the significant improvement in the accuracy ~of_the critical. heat flux predictions over previous DNB correlations. . i Increased Rod Drop Time The V5B fuel design incorporates a snag resistant, low-pressure-drop sircaloy grid. The sircaloy grid will pre :te for an enchanced performance relative to the current 17 x : 7
?TD fuel used at SQN.
I Utilisation of aircaloy as a grid material zustead of Inconel reduces ! the source of cobalt in the core. Consequently, radiation fields caused by the transport of activated cobalt should be lower. The snag-resistant feature results from outer grid straps that are ; modified to reduce'the potential for grid damage and assembly' hang-up i from assembly interactions during fuel assembly removal. The , zircaloy grid also contains features that minimize hydraulic ! resistance. ! In order to maintain mechanical compatibility between the V5H grid , and guide tube, a reduction in the V5H guide tube diameter was i required. The allowable rod-drop time of TS 3.1.3.4 must be increased because of the increased dashpot effect resulting from the guide tube diameter reduction. Elimination of Rod Bow Penalty Fuel rod bow has been observed in Westinghouse cores. 'The r.enomene , of fuel rod bowing must be accounted for in the DNB safety analyses
.of Condition I and II events. The current licensing basis offsets the DNB effects of rod bow by partially accommodating it with margin in the W-3 correlation. The remainder of the rod bow penalty is appliedasapenaltyontheF{Hevaluation.
New statistical methods have been developed by Westinghouse that verify that the past treatment of rod bow penalty provided an overestimation of the effects on DNB. Application of the new methods to SQN for the standard and the V5H fuel products has verified the reduction in rod bow penalty. The reduction allows for accommodation of the entire penalty in the establishnent of the safety limit DNBR. I i (
[{ , 1 b i - 3 g., m
<h
"~
[" 1 The; requested change to SQN TS 3.2.3 will remove an unnecessary power
' peaking' penalty and simplify the format of the specification. The F current specification format includes RCS flow not only to maintain '
>the-minimum required RCS flow but also to provide for an additional offset against rod bow penalty;- The use of RCS flow to compensate-for rod bow penalty will no longer be required.-
Theflowlimit(anditsassocia$eduncertaintyfactor)asitrelates. ' O
.to DNA will be moved to TS 3.2.5. DNB parameters, as discussed ,
below. Similar TS changes related to rod bow penalty have been performed for Farley Units 1 and 2, North Anna Units 1 and 2, Beaver Valley Unit 1, and Salem Unit.2.
^
Definition'of DNB Parameter RCS Flow Limit The RCS flow limit and its associated uncertainty factor have been moved-to TS 3.2.5. DNB parameters, which now establishes a minimum allowable RCS flow to prevent violation ~of the safety limit DNB' during normal operation'and accident conditions. The minimum flow rate is based on a thermal design flow rate of
-365,600 gym plus the application of a correction for measurement uncertainties'(minimum flow rate = 365,600 gym x uncertainty factor). The uncertainty factor of 3.5 percent is based on flow-
. measurement uncertainties and feedweter venturi fouling. Therefore, ,
the RCS flow limit for SQN's units is 365,600 spm x 1.035, or , 378,400 gpm. ' Environmental Impact Evaluation The proposed change request does not involve an unreviewed environmental
-question because operation of SQN Units'1 and 2 in accordance with this n: change would not:
-1. Result in a significant increase ~in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by.the Staff's testimony to the Atomic Safety and Licensing Board, supplements to the.FES, environmental impact appraisals, or-decisions of the' Atomic Safety and Licensing Board.
- 2. Result in a significant change in effluents or power levels.
d
.I
- 3. Result in matters not previously reviewed in the licensing basis for l SQN that may have a significant environmental impact.
1 i I l
s [. l , p , 4 L l ,i,
, t 't
['.
'j(. . " -i l
l ENCLOSURE 3.
+ '
PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 I (TVA-SQN-Iti-89-33) 1 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS-FOR VANTAGE 5 HYBRID FUEL UPGRADE ' 5 i P. I [5 --
.,e- .u, 4 4 e h
I 9
- l l
l L l l 1 I 1 1
P F k (' 4 I i l ENCLOSURE 3 1 Significant Hazards Evaluation TVA has evaluated the proposed TS change and has determined that it does not represent aLsignificant hasards consideration based on criteria established in 10 CFR 50.92(c). Operation of SQN in accordance with the proposed amendment will nott (1) ' Involve a significant increase in the probability or consequences of an accident previously evaluated.
-The evaluations of the mechanical, nuclear, and thermal-hydraulic v .
design effects support the. conclusion that the requested changes are within the current design criteria established in the FSAR. Consequently, no new mechanisms have been introduced to increase the probability of a previously analyzed accident occurring. The accident evaluations (both LOCA and non-LOCA) exhibit results that maintain the confidence level in the physical integrity of the fission product boundaries as defined in the FSAR. Therefore', the consequences of the accidents do not increase. (2) Create the possibility of a new or different kind of accident from ' any previously analyzed. The evaluations performed established that the FSAR design criteria and system responsee during normal and accident conditions are bounding with respect to the proposed changes. The changes will not affect.the function of any protection system, and they will not introduce hardware that is different in design criteria requirements. - Therefore, no new mechanisms have been introduced that would create the possibility of a new or different kind of accident from those previously analyzed. (3) Involve a significant reduction in a margin of safety. The evaluations performed addressed all design criteria and accident analyses. In performing the evaluations, the safety limits , established by the FSAR and TS were not modified such as to reduce i the difference between the safety limit and the limit defined as the failure point of a fission product boundary. Therefore, the margins that were assumed in the accident analyses remain bounding for the proposed changes.
p -:7 ,
.3 L ;'l > l
.y. ! , i - t . _ - ) _f .
- t. -- 1
'r.., 's si
. y i.:
' . ENCLOSURE 4 l + PLANT: SAFETY EVALUATION FOR SEQUOYAH NUCLEAR PLANT n'; UNITS 1 AND 2 L , VANTAGE 5 HYBRID FUEL' UPGRADE fi' I i
\
1 g e
'4 t
4 9 e O
; _1..
-; g IN ', , 1 j i ' ,
. h* n : i g* '
.t
..w
, p
.i i
- n. .
?
PLANT SAFETY EVALUATION FOR- t w SEQUOYAH ' NUCLEAR PLANT .- Py ' q UNITS 1 AND 2
. VANTAGE 5H-FUEL UPGRADE i
NOVEMBER 1989 Editors: B. W. Gergos . l . L V. Tomasic Contributors:. F. Baskerville P. Kersting R. Brashier . . J. Killimayer C. Brockhoff N. Pogorzelski J. Grover L Schaub
. D. Kelly W. Schiviey L R. Kemper L
a r Approved: E. H. Novenstern, Manager T/H Design & Fuel Licensing ll WESTINGHOUSE ELECIRIC CORPORATION ! Commercial Nuclear Fuel Division i P. O. Box 3912 Pittsburgh, Pennsylvania 15230 Attachment to THFL-89499
+ '9- T " .
r , ., ., . . _ .
2 n. M > p<
~
PLANT SAFEIY EVALUATION - FOR THE, i
. SEQUOYAH NUCLEAR PIANTS UNITS 1 AND 2 FUEL UPGRADE -
7 TABLE OF CONTENTS SECTION IIILE EKE s 1.01 INTRODUCTION AND
SUMMARY
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 1-1 ;
4 h 2.0 : DESIGN FEATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 221 2.1 - Introd uction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.2 VANTAGE SH Fuel Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1: . L 2.3 VANTAGE SH Zircaloy Grit . . . . . . . . . . . . . . . . . . . . . . . . . . . -. . . . . . . 21
. 2.4 Fuel Rod Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 b,
3.0 . NUCLEAR DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.1 Introduction and Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 ! 3.2 Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . - . . . . . . . . . . . . . . . . . . 3-1 3.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 t 3.4 Conclusion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
~ 4.0 THERMAL AND HYDRAULIC DESIGN . . . . . . . . . . . . . . . . . . . . . . : . . . . . . 4-1
'4.1 Introduction and Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 C 4.2 Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.3 Hydraulic Compatibility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.4 . Effects of Fuel Rod Bow en DNBR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.5 Fuel Temperature Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 4.6 ' Transition Core Effect . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.7 - Conclusion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 1 5.0 - ACCIDENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 L 5.1 Non-LOCA Accidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 5.2 LOCA Accidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 25 5.3 Environmental Consequences of Accidents ....................... 5-28 1-l 6.0 REFEREN CES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 l
APPENDIX A Summary of Technical Specification Changes L- '
' APPENDIX B L Significant Hazards Evaluation 1
1 APPENDIX C Nuclear Safety Evaluation Checklist l l i5 l
x N
%v.
'.f^i.t< -~j . . , W
'JW ,
LIST OF MGURES
.(.
/ -
~HQ.UBE TITLE PAGE
. 1
' 21 '
l ~ Co.mparison of the 17X17. VANTAGE SH Fuel Assembly
. and the 17X17 STD Fuel Assembly . . . . . . . . . . . . . .. . . .. . . . . . . . . . . . 23' 1
. 5.1 1 Startup from Subcritical
'y" ,
Nuclear Power versus Time ................................ 5 11 5.1 2 - Startup from Subcritical
; Heat Flux versus Tune . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-12 v.
p - 5.1-3' Uncontrolled Rod Withdrawal from a Subcritical Condition .
~
- Fuel and Clad Temperature versus Time ....................... 5 13
. 5.1-4 Partial loss of Forced Reactor Coolant Flow Core and loop Flow versu: Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 5 14 5.1 5 Partial Loss of Forced Reactor Coolant Flow Heat Flux versus Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 15 s I
l 5.14 PartialIoss of Forced Resctor Coolant Flow i t DNBR versus Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 16 ) i 5.1 7 Complete Loss of Forced Reactor Coolant Flow l 2
~ Core and Loop Flow versus Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 17 1 1
5.1 8 Complete Imss of Forced Reactor Coolant Flow l Heat Flux versus Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-18 ! 5.19 Complete Loss of Forced Reactor Coolant Flow i DNBR versus Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-19 p 5.1 10 'Startup of an Inactive loop L Heat Flux versus Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-20 5.1-11 Startup of an Inactive loop DNBR versus Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 21 l
- l. 5.1-12 Single Reactor Coolant Pump Locked Rotor
! Core Flow versus Time ................................... 5 22 / , L 5.1-13 Single Reactor Coolant Pump Locked Rotor l Nuclear Power and Heat Flux versus Time . . . . . . . . . . . . . . . . . . . . . . 5 23 i L 5.1-14 Single Reactor Coolant Pump Locked Rotor l Clad Temperature versus Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 24 1 1 I 11 I
- . _ . _ m._.m .____-__.-_m____._.____._..__ -
~ - . .. -
i J ai-e, ' , i' 1 f LIST OF TABLES s ( TABLE TITLE E60E
-l 1
l 2-1 Comparison of 17X17 Standard and VANTAGE 5 Hybrid Fuel Assembly . . . . . . . . . . . . . . . . . . . . . . . . 2-4 il 41- 'Ibermal and Hydraulic Design Parameters . . . . . . . . . . . . . . , . . , , . . . 44 :
' 4-2 -- DNB Margin Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , , . 4 6
'5.1 1' Rod Cluster Control Assembly Ejection
. . Accident ' Results . . :. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , , , , , 39 <
5.1 2 Summary of Results for Locked Rotor Transients . . . . . . . . . . . . . . . . . 5 10 l l b t[ 111
g ;; , , y l' m -
)
1 y A j 1.0 INTRODUCrlON AND
SUMMARY
1 4 Ci, . H It.is planned to refuel and operate the Sequoyah Nuclear Plant Units 1 and 2 with Westinghouse fuel employing advanced fuel product features. As' a result, future core loadings will have fuel - assemblies that incorporate' a low pressure dr'op Zircaloy grid, Removable Top Nozzles, Integral Fuel Burnable Absorber and extended burnup capability. His upgraded fuel feature is known as - VANTAGE 5 Hybrid (VANTAGE SH) and has been submitted as Addendum 2 A (Reference 1) l to the " VANTAGE 5 Reference Core Report,' WCAP 10444 P-A (Reference 2). In addition to
)
the above features, future Sequoyah reloads will also contain Debris Filter Bottom Nozzles, snag resistant ' grids and standardized pellets; these additional features have been implemented in other Westinghouse reload cores and do not require prior NRC approval since safety evaluations have shown that a 10CFR50.59 determination can be made for these design features. A brief summary of'the upgraded fuel features are given below. VANTAGE SH Assembly - The VANTAGE SH fuel assembly design evolved from the current VANTAGE 5, Optimized (OFA) and Standard (STD) fuel assembly designs. It is based on substantial design and operating experience. Design features from each of these previous designs ( are incorporated into the VANTAGE SH fuel assembly design. De most significant VANTAGE. 5H design change is the use of Zircaloy grids with 0.374 inch OD standard fuel rods. To l accommodate the.Zire grids, the VANTAGE SH thimble tube diameter was modified to be the same as the 17x17 OFA or VANTAGE 5 fuel. Removable Ton Nozzle (RTN) - He RTN differs from the current design in two ways: a groove
. is provided in each thimble thru hole in the nozzle plate to facilitate attachment and removal; and the nozzle plate thickness was reduced to provide additional space for fuel rod growth. In conjunction with the RTN,'a long tapered fuel rod bottom end plug is used to facilitate removal and reinsertion of the fuel rods.
w Intecral Fuel Burnable Absorber (IFBA) - De IFBA features a zirconium diboride coating on the fuel pellet surface on the central portion of the enriched UO2 fuel stack. In a typical reload core, approximately one third of the fuel rods in the feed region are expected to include IFBAs. IFBAs provide power peaking and moderator temperature coefficient control. 1-1 1
by' - y l
?
a. n 1 Extended Burnun - ne VANTAGE SH fuel design will be capable of achieving extended burnups. C4- The basis for designing to extended burnup is contained in Reference 3.
. 1
- Debris Filter Bottom' Nozzle (DFBN) . His bottom nozzle is designed to inhibit debris from -
entering the active fuel region of the core and thereby improves fuel performance by precluding
. debris related fuel failures. The DFBN is's low profile bottom nozzle design made of stainless - t steel, with reduced plate thickness and leg height to provide additional space for fuel rod growth.
The DFBN is structurally and hydraulically equivalent to the existing standard bottom nozzle. Snac-Re'sistant Grids The snag resistant grids contain outer grid straps which are modified to ~ > prevent assembly hangup from grid strap interference during fuel assembly removal. This was accomplished by changing the grid strap corner geometry and the addition of guide tabs on the outer grid strap. Standardized Fuel Pellet - The standardized pellet is a refinement to the current pellet designs - with the objective of improving manufacturability while maintaining or improving performance. This design incorporates a reduced pellet length and modification to the previous dish size. The chamfer feature which was included in the pellets for Sequoyah Unit 2 will be introduced for l Sequoyah Unit 1. The Sequoyah Units 1 and 2 Plant Safety Evaluation (PSE) is to serve as a reference safety li evaluation / analysis report for the region by-region reload transition from the present Sequoyah - Unit I and Unit'2 cores to a core containing the above upgraded features. 1 The PSE utilizes the standard reload design methods described in Reference 4 and will be used
- as a basic reference document in support of future Sequoyah Units 1 and 2 Reload Safety
~ Evaluations (RSEs) for upgraded fuel reloads. Sections 2.0 through 5.0 of the PSE provides the l
12 1
g; y , ,
" ~
results of the Mechanical, Nuclear, Dermal 'and Hydraulic and the Accident Evaluations. I
' Appendix 'A gives' a summary of the changes to the Technical Specifications required and the ,
k corresponding change pages. De Significant Hazards Evaluation and the Nuclear Safety Evaluation Checklist are provided in Appendices B and C respectively. Consistent with the Westinghouse standard reload methodology, Reference 4, parameters are l chosen to maximize the applicability of the. PSE evaluations 'for future cycles. The objective of $ subsequent cycle specific RSEs will be to verify that applicable safety limits are satisfied based on the reference evaluation / analyses established in this safety evaluation. l The results of evaluation / analysis described herein lead to the following conclusions: l l
- 1. The Westingbouse fuel assemblies containing VANTAGE SH and the additional (upgraded fuel features for the Sequoyah Units are mechanically compatible with l 9
't he current fuel assemblies, control rod and reactor internals interfaces. The current a design bases for Sequoyah Units 1 and 2 have been changed as described in this report to accommodate the VANTAGE SH design.
- 2. Changes in the nuclear characteristics due to the transition to upgraded fuel will be within the range normally seen from cycle to cycle due to fuel management effects.
- 3. The upgraded reload fuel assemblies a*e ; hydraulically compatible with the fuel assemblies from previous reload cores.
- 4. The core design and safety analyses results documented in this report show the core's capability for operating safely for the rated Sequoyah design thermal power.
- 5. Previously reviewed and licensed safety limits are met when the Sequoyah Units are reloaded with upgraded fuel as described in this report. Plant operating C
1-3
t... , x ,. , ,, f ",i , ' -) /,' %' , I.' ' ',, j ld[,i % <! , n} i *, ,
,i l !- ~! 'F
-.; s . %
.lgi "
- 3 p!;i"le, f;::;..-
- rA *
- v. ~t- 6 f I t ' h b 1"' -t r <
i6 - ' - '.f'% '>- ,, ,
, j ll4., ;,3 '
NT: , limitations given in the Technical Specifications will be satisfied with the proposed to@ . . . pA ,
- changes noted in Appendix A of this report. A reference has bcen established upon -
])
- which toLbase Westinghouse reload safety evaluations for future reloads with the~
ll? w ' .u.' ,, ,
,_ .1 l) . , upgraded fuel features. : :- @ I L
I.f. . . _ 1 r-3 Il
)
- p. :
E. I F {
?
eq n
.{
o e
)
4 k
-?) .
3 l '- i i l, l.. l 1-4 os . ._ _
y L , O p I; 2.0 DESIGN FEATURES
, C ~2 .1 c Introduction n i l The mechanical design of the upgraded fuel' assemblies for Sequoyah Units 1 and 2 is the same as previous reload fuel assemblies except the upgraded fuel assemblies will incorporate several fuel design improvements. These improvements include the VANTAGE SH design features' described
. in Section 1 with the addition of Debris Filter Bottom Nozzles, snag resistant grids and standardized fuel pellets.
2.2 - VANTAGE SH Fuel Assembly 4 The primary mechanical difference between the 17x17 VANTAGE SH design relative to the 17x17 STD fuel currently in the Sequoyah Units is the use of the VANTAGE SH Zircaloy grid. A
; comparison of the STD and VANTAGE SH fuel assembly design parameters is given in Table 2-1. Figure 21 demonstrates the similarity of the two designs and shows a comparison of overall dimensions.
Comparative fuel assembly flow testing pressure drop results indicate that the VANTAGE SH and the STD 17x17 fuel assembly are hydraulically equivalent. Full assembly testing has confirmed that the VANTAGE SH fuel assembly has hydraulic stability and that the fuel rod contact wear with the spacer grids is within the allowable design limits. The major components that determine the structural integrity of the fuel assembly are the grids. Mechanical testing and analysis of the VANTAGE SH Zircaloy grid and fuel assembly have demonstrated that the VANTAGE SH structural integrity under seismic /LOCA loads will provide margins comparable to the STD 17x17 fuel assembly design and will meet all design bases. 2.3 VANTAGE SH Zircaloy Grid The VANTAGE SH Zircaloy grid is based on the OFA Zircaloy grid design and operating experience. The grid strip thickness, type of strap welding, basic mixing vane design and pattern, method of thimble tube attachment, type of fuel rod support (6 point), material and envelope are identical to the OFA Zircaloy grid. 21
v n- 1 i[ . 1 _' 1 , z go F' ' In order to demonstrate early performance of the Zircaloy grid design, fuel assembly demonstration programs were conducted inserting OFA fuel assemblies containing Zircaloy grids into 14A14,15x15 l and 17x17 cores. Subsequent to the satisfactory performances observed in these programs, the OFA . with Zircaloy grids were loaded and-have operated successfully since the early 1980's in many l 3 ' Westinghouse cores. His experience is documented in Reference 5. Relative to the OFA grid, the VANTAGE SH grid includes features that minimize hydraulic resistance while maintaining 1 required structural capability. Ris evaluation of the VANTAGE SH grid performance is based on the extensive design _and' irradiation experience with previous grid designs and full grid testing I completed with the VANTAGE SH grid design. I 2.4 Fuel Rod Performance he 0.374 inch OD fuel rod used in the VANTAGE SH fuel assembly is the same as that used in the Sequoyah 17x17 STD fuel assemblies. The design basis, methodology, and models are the , same as those previously described in References 2' and 3 with the exception that the NRC approved Westinghouse fuel performance models in Reference 6 are being used. No changes in ( fuel rod design criteria, methods or models are necessary because of the transition to VANTAGE SH fuel.
.I 1..
i 22 l~
g '
, fl -
, 's-j l FIGURE 2-1 Comparison of the 17X17 VANTAGE SH Fuel Assembly and the 17X17 STD Fuel Assembly ,
'li : , 17X17 VANTAGE SH FUEL ASSEMBLY
= . . . . = ,.
= = 3.47 2.343 *
= -
152.2 t P ' l l a !!,I g t li R ! E II H il l t l 17XI7 RECONSTITUTABLE STD FUEL ASSEMBLY fv Z q 199.768 , 2 3.670 . 3.73s + e j 151.96 : x 9IlilillIIIIII!!K 2-3
e t. 0 1:. ,, , m , n--- . TABLE 21 1 Comparison of 17X17 Standard ' and VANTAGE 5 Hybrid Fuel Assembly .. Mechanical Design Parameters F 2' Standard VANTAGE SH Fuel Assembly Overall I.ength, inch 159.8 160.0 Fuel Rod Overall Length, inch 151.6 152.2 ' t Assembly Envelope, inches. 8.426 8.426 Fuel Rod Pitch, inch 0.4 % 0.4% . l
~ Number of Fuel Rods / Assembly 264 264 Number of Guide Thimbles / 24 24 ,
- Assembly .
Number of Instrumentation 1 1 Tube / Assembly -; ( Fuel Tube Material
' Fuel Tube Clad OD, inch Zirc4 Zirc-4
' l
{ 0374 0374 - h Fuel Rod Clad 'Ibickness, inch 0.0225 0.0225 . t i Fuel Clad Gap, mil- 6.5 - 6.5 l (uncoated ., pellets) e I i Fuel Pellet Diameter, inch 03225 03225 ] (uncoated i pe!!ets) ! Fue! Rod End Plugs Standard Tapered Fuel Pellet I.ength . Enriched fuel, inch 0387 0387 , Unenriched fuel, inch NA 0.545 1 i
?
24
i
.c i y !
4 TABI.E 21 (Cont)
- Comparison of 17X17 Standard L and VANTAGE 5 Hybrid Fuel Assembly .
i N Mechanical Design Parameten U k Standard YAFTAGESH Relative Qad Dicknesst 1.0 1.0 _ Diameter Ratio Relative Moderator / Fuel . . 1.0 1.0 Ratio for Assembly ; 1 < e' -Relative UO 2/ Rod 1.0 1.0
= Ouide Thimble Material Zirc 4 Zirc-4 Guide *!himble OD, inch 0.482 0.474 i ;
W Oulde Thimble Wall 0.016 0.016 nickness, inch Orid Material Inner Inconel Zirc-4 Mid Grid (6) { Edgm Modified No Yes ! t Grid Material, End Orids(2);
- Inconel Inconel )
Orid Types Utilized Inconel Mid Grids- Yes No Zircaloy Mid Grids No Yes L Inconel Top & Bottom Orids Yes Yes Inner Spring (Mid Grids) Vertical Non Vertical
- 1 ;
4 g I 25 l'
I. r, i-TABLE 21 (Cont) , Comp.trison of 17X17 Standard o and VANTAGE $ Hybrid Fuel Assembly . Mechanical Design Parameters i i Standard VANTAGE $H Grid Fabrication Brazed joining Brazed joining Inconel Grids of interlocking of interlesking stamped straps stamped straps p Zarealoy Mid Grid None Laser weld joining of interlocking stamped straps ', Grid / Guide nimble Attach. nimbles bu%ed nimbles bulged Inconel Grids together with together with sleeves prebrazed sleeves prebrazed Grid / Guide nimble Attach. C Zircaloy Mid Grids None nimbles bulged together with sleeves laser prewelded to straps Top Nozzle Welded stainless Removable steel Standard stdaless steel reduced height removable design Compatible with Fuel Yes Yes Handling Equipment t 26
l t 3.0 NUCIEAR DESIGN l C 3.1 Introduction and Summary I This section provides an assessment of the nuclear design impact from the upgrade in fuel product , from 17x17 STD to 17x17 VANTAGE SH for the Sequoyah units. De fuel design changes t impacting nuclear design result from the incorporation of Zircaloy grids, IFBA and increased > discharge burnup. The effect of these changer on the core physics characteristics are small and explicitly modeled in the neutronics codes. De specific values of core safety parameters, e.g. power distributions, peaking factors, rod worths, are primarily loading pattern dependent. The variations in the loading pattern dependent safety parameters are also expected to be typical of the normal cycle to cycle variations for the standard fuel reloads. No change in Technical Specifications related to core neutronics behavior result as a consequence of the upgrade in fuel product. In summary, the change from the current all standard fuel core to a core containing the upgraded fuel product will not cause changes to the current Sequoyah FSAR nuclear design bases. 3.2 Methodology No changes to the nuclear design philosophy or methods are necessary because of the upgraded i fuel product. De reload design philosophy includes the evaluation of the reload core key safety parameters which comprise the nuclear design dependent input to the FSAR safety evaluation for each reload cycle. his philosophy is described in Reference 4. These key safety parameters will be evalueted for each Sequoyah reload cycle. If one or more of the parameters fall outside the bounds assumed in the safety analysis, the affected transient will be re-evaluated and the results documented in the RSE for that cycle. 3.3 Design Evaluation , ne 0.374 inch diameter fuel rod has had extensive nuclear design and operating experience with the current Sequoyah 17x17 STD fuel assembly design. De Zircaloy grid material has also had extensive nuclear design and operating experience with the current 17x17 VANTAGE 5 and 17x17 l 31
t ( i OFA fuel assembly designs. IFBAs are used for power distribution control similar to fixed burnable absorbers. De actual peaking factor characteristics are loading pattern dependent. IFBAs may { be used with fixed burnable absorbers in the same assembly to increase design flexibility. , l 3.4 Conclusion i he VANTAGE SH and SID assembly are neutronically similar. Any variation in design is f expected to be within the cycle to cycle variations for the STD core. The nuclear design philosophy and process is unchanged due to the VANTAGE SH fuel. i e 32
( l ( 4.0 THERMAL AND HYDRAULIC DESIGN ' N 4.1 Introduction and Summary His section describes the calculational methods used for the thermal-hydraulic analysis, the DNB perforinance and the hydraulic compatibility during the transition from mixed fuel cores to an all VANTAGE 5H core. Based on minimal hardware design differences and prototype hydraulic { testing of the fuel assemblies, it is concluded in Reference 1 that the STD and VANTAGE SH I fuel assembly designs are hydraulically compatible. Table 41 summarizes the thermal hydraulic design parameters for Sequoyah Units 1 and 2 that were used in this analysis. In addition, the y current rod bow methodology is implemented to reduce the rod bow penalty described in Sequoyah Technical Specifications. He thermal hydraulic design criteria and methods remain the same as those presented in the Sequoyah FSAR with the exception noted in the following sections. All of the current FSAR thermal hydraulic design criteria are satisfied. 4.2 Methodology Calculational methods currently used on the 17x17 STD fuel assembly and described in the Sequoyah FSAR are applicable to the evaluation of the core containing both 17x17 STD and VANTAGE SH fuel assemblies, except for the DNB correlation. The analyses for STD and VANTAGE SH fuel will utilize the WRB 1 DNB correlation (Reference 7). The WRB 1 DNB correlation is based ~ entirely on rod bundle data and takes credit for the significant improvement in the accuracy of the critical heat flux predictions over previous DNB correlations. As documented in Reference 7, a 9595 limit DNBR of 1.17 is appropriate for the STD fuel assemblies. ne WRB 1 DNB correlation is applicable to VANTAGE SH fuel since, from a DNB perspective, the VANTAGE SH assembly is virtually identical to the 17x17 STD design. As documented in Reference 1, the use of the WRB 1 DNB correlation with a 95SS limit DNBR of 1.17 is applicable to the VANTAGE SH fuel assembly. DNBR margin is maintained for the STD and VANTAGE SH fuel by performing the DNB safety analysis to a DNBR limit of 1.38. Comparing this limit of 1.38 to the WRB 1 correlation limit of 41
u
, 1.17 results in 15.2% DNBR margin. his DNBR margin is more than sufficient to offset the 5A DN3R penalty due to rod bow (see Section 4.4).
(' Table 4 2 summarizes the DNBR margin and the allocation of margin for the Sequoyah plants. DNBR margin in excess of the allocation to cover DNBR penalties is available for plant design . . flexibility, c. 4.3 Hydraulic Compatibility The STD fuel assembly and VANTAGE SH design have been shown in Reference 1 to be ". o hydraulically compatible. _2 4.4 Effects of Fuel Rod Bow on DNBR i ! De phenomenon of fuel rod bowing must be accounted for in the DNBR safety analysis of Condition I and Condition U events. Currently the rod bow penalty is partially offset by the . N generic DNBR margin. De remaining rod bow penalty is applied as agnalty on F AHas { described in the Sequoyah Technical Specifications, which defines the FAH as a function of region ' - average burnup to 33,000 MWD /MTU, For this application the rod bow penalty was reassessed $ using improved methodology. Using the current methodology described in References 8,9 and 10, a rod bow penalty of <1.5% t is applicable to 17x17 STD fuel assemblies. Based on the similarities between 17x17 STD and ; VANTAGE 5H fuel assemblies, (i.e. fuel rod diameter, fuel rod pitch and grid spacing), this penalty is also applicable to VANTAGE SH fuel assembly. , f Dis penalty is the maximum rod bow penalty at an assembly average burnup of 24,000 MWD /MTU. For burnups greater than 24,000 MWD /MTU, credit is taken for the effect of FfH burndown, due to the decrease in fissionable isotopes and the buildup of fission product inventory. Herefore, no additional rod bow penalty is required at burnups greater than 24,000 MWD /MTU. [ci, 42 4
i
)
4.5 Fuel Temperature Analysis C he fuel temperatures for use in safety calculations for the VANTAGE SH fuel are identical to , ) those for the STD fuel.
'4.6 Transition Core Effect i
ne VANTAGE SH hyoraulic test program showed identical results for the VANTAGE SH ; Zircaloy mixing vane grid and the STD fuel Inconel mixing vane grid. Derefore, no transition core DNBR penalty is necessary. ' 4.7 Conclusion # r The thermal. hydraulic analysis has shown that 17x17 STD and VANTAGE SH fuel assemblies are hydraulically compatible and sufficient DNBR margin in the safety limit DNBR exists to cover the [ rod bow penalty. All current thermal hydraulic design criteria are satisfied. 9 9 t l C 43
I i l L TABLE 41 Thermal and Hydraulic Design Parameters Thermal and Hvdraulie Desien Parameters Desien Parameten , i Reactor Core Heat Output, MWt 3411 ! Reactor Core Heat Output,106 BTU /hr 11,642 Heat Generated in Fuel, % 97.4 Core Pressure, Nominal, psia 2250 Radial Power Distribution 1.55[1 +0.3(1.P)) Limit DNBR for Design Transients (STD) 1.38 (V5H) 1.38 . , DNB Correlation (STD) WRB.1 (V5H) WRB.1 > HFP Nominal Coolant Conditions , Vessel Thermal Design Flow Rate (including Bypass), 106 lbm/hr 138.0 GPM 365,600 . Core Flow Rate * (excluding Bypass, based on TDF) l 106 lbm/hr 127.7 GPM 338,180 Fuel Assembly Flow Area for Heat Transfer, ft2 (STD) 51.1 (V5H) 51.3 Core Inlet Mass Velocity, 6 10 lb:n/hr-ft2 (Based on TDF) (STD) 2.50 (V5H) 2.49
- Based on design bypass flow of 7.5%.
C
~
I TABLE 41 (Continued)
*Ihermal and Hydraulic Design Parameters hrmal and Hvdrate Lwl-n Parameters Desien Parameters ;
Nominal VessevCore Inlet Temperature, OF $46.7* Vessel Average Temperature, OF $78.2 Core Average Temperature, OF , 582.2 Vessel Outlet Temperature, OF 609.7 Average Temperature Rise in Vessel, OF 63.0 Average Temperature Rise in Core, OF 67.6 - Heat Transfer . Active Heat Transfer Surface Area, ft2 (STD) 59,700 (V5H) 59,700 Average Heat Flux, BTU /hr ft2 (STD) 189,800 (V5H) 189,800 Average Linear Power, kw/ft 5.44 Peak Linear Power for Normal Operation,+ kw/ft 12.6 Temperature at Peak Linear Power for Prevention of Centerline Melt, OF 4700
- Safety analysis inlet temperature is $48.20F
+ Based on 2.32 Fo peaking factor 45
i TABLE 4 2 DNB Margin Summary ; DNB Correlation WRB1 Correlation Umit 1.17 i Safety Limit 1.38 , DNBR Margin + 15.2 % [ DNBR Penalties ; Rod Bow <1.5% ' Transition Core 0.0%
- r i
t
+ DNBR margin between the safety limit DNBR and the correlation limit DNBR.
P 4-6
i b 1 5.0 ACCIDENT ANAIXSIS 5.1 Non-LOCA Accidents His section summarizes the reanalysis and evaluations perfonned for the Sequoyah Units upgrade to VANTAGE SH fuel. J he major effect of changing from STD 17x17 fuel to VANTAGE SH fuel on the non.LOCA . transients is the increased Rod Control Cluster Assembly (RCCA) drop time. De VANTAGE 5H fuel assembly has a thimble tube I.D. of 0.442 inches. STD fuel has a thimble tube I.D. of 0.450 inches. De smaller VANTAGE SH thimble tube will increase the RCCA drop time from a current limit of 2.2 seconds to 2.7 seconds. This slower drop time will affect the results of the fast non.LOCA limiting transients such as Partial and Complete 1.oss of Forced Reactor Coolant Flow, RCCA Bank Withdrawal from Suberitical, Single Reactor Coolant Pump I.ocked Rotor and Rod Ejection. These transients were reanalyzed and a summary of the results of the analyses are - discussed on the following pages. In addition, Startup of an Inactive I. cop was analyzed because of the change in the DNBR correlation. Non.LOCA events not mentioned above were not reanalyzed for the increased rod insertion time - for one or more of the following reasons:
- 1) Transient ruults are insensitive to the rod insertion rate.
- 2) Reactor trip was not assumed or explicitly modeled in the analysis.
- 3) Reactor trip has no effect on the minimu n or maximum value of the critical parameter of interest.
- 4) Event may be impacted, but the magnitude of the impact is small when compared to the margin to the design basis limit.
The non.LOCA analyses and evaluations for VANTAGE SH were performed taking into account the effects from the following programs: Eagle 21 Digital Protection System Resistance Temperature Detector (RTD) Bypass Elimination { - New Steamline Break Protection 51 4
l I f I.ow Feedwater Flow Reactor Trip Elimination l Boron Injection Tank (BIT) Removal ! Upper Head Injection (UHI) Removal 1 Uncontrolled RCCA Withdrawal from a Suberitical Condkion (FSAR Section 15.2.1) i ne RCCA withdrawal from suberitical is characterized by a rapid power increase. The power excursion is retarded by Doppler feedback and the transient is terminated by a reactor trip. Due to the high rate at which the power increases, this transient can be sensitive to RCCA drop time. ' a
. The RCCA withdrawal from suberitical was reanalyzed with a RCCA drop time to the dashpot of 2.7 seconds. Transient results are shown in Figures 5.11 through 5.13. Figure 5.11 shows the neutron flux transient. The neutron flux overshoots the full power nomir al value, but this occurs for only a very short time period. De energy release and the fuel temperature increases are [
relatively small. ne thermal flux response, of interest for DNB considerations, is shown on Figure 5.1-2. De beneficial effect on the inherent thermal lag in the fuel is evidenced by a peak heat flux less than the full power nominal value. There is a large margin to DNB during the transient ' since the rod surface heat flux remains below the design value, and there is a high degree of
'( subcooling at all times in the core. Figure 5.13 shows the response of the average fuel and cladding temperature. ne average fuel temperature increases to a value lower than the nominal full power value, ne minimum DNBR at all times remains above the limits. -
The core and the Reactor Coolant System (RCS) are not adversely affected, since the combination - of thermal power and the coolant temperature result in a minimum DNBR well above the limiting i value. Dus no fuel or clad damage will occur, i Uncontrolled RCCA Bank Withdrawal at Power (FSAR Section 15.2.2) This event is analyzed to show that the DNB design basis is met. De Overtemperature AT setpoints are not changed, so the time of reactor trip remains the same. The transient minimum DNBR occurs immediately following the reactor trip. A conservative evaluation of the effects of the 0.5 second increase in control rod drop time was done by extrapolating the transient DNBR results assuming that the reactor trip was delayed by 0.5 second. De extrapolations showed that ample mugin to the DNB limit still exists with a 0.5 second delay. The conclusions of the FSAR remain valid. 52
, e - _ _ - - - - - - . - - - -
i l Rod Clus.tcLControl Assembh> Misalinnment (FSAR Section 15.2.3) Dese events are analyzed to show that the DNB design basis is met, ne dropped rod analysis was updated for VANTAGE SH fuel and the increased RCCA drop time using the current Westinghouse methodology. For a 2.7 second drop time, the maximum rod worth which is ! considered in the analysis is 500 pcm. For rod worths above this limit, a reactor trip is assumed to occur and no further analysis is needed. The negative Dux rate trip protection system might not, however, detect rod worths less than 500 pcm. %erefore analyses are performed each cycle to l ensure that the DNB design limit is met for dropped rod worths less than 500 pcm. ne Statically l h'isaligned RCCA events are not impacted by the increase in rod drop time. Dus, the analysis results are unaffected and the conclusions of the FSAR remain valid. - Uncontrolled Boron Dilution (FSAR Section 15.2.4) Acceptable results for this event are demonstrated by showing that there is sufficient time for i operator action to terminate a dilution prior to a return to criticality, which could lead to core damage. For the case analyzed at power with manual rod control, the time of trip is determined by comparison to the RCCA bank withdrawal at power analysis. As previously noted, the time ( of reactor trip on Overtemperature AT is unchanged. The mechanics of the reactor trip are not explicitly modeled in the analysis. Rus, the analpis results are unaffected by the increased rod insertion time and the conclusions of the FSAR remain valid. Partial and Comnlete Lee:S of Forced Reactor Coolant Flow (FSAR Section 15.2.5 & 15.3.4.1 he loss of flow transients are characterized by a rapid decrease in core Dow. If the reactor is not tripped promptly DNB may occur with subsequent fuel damage. Thus these transients can be sensitive to the RCCA drop time. De partial and complete loss of Dow transients were reanalyzed with a RCCA drop time (time to dashpot) of 2.7 seconds. Figures 5.1-4 through 5.1-6 show the results of the partial loss of flow. He results of the complete loss of Dow transient are shown in Figures 5.1-7 through 5.19. In both cases the minimum DNBR remains above the limit. The ' underfrequency Loss of Flow transient was also analyzed and the result was that the minimum DNBR remains above the limit. .
.Startuo of an Inactive Reactor Coolant Loon (FSAR Section 15.2.61 This event is analyzed to show that the DNB design basis is met. Minimum DNBR occurs immediately following a reactor trip on high neutron Dux. This transient was reanalyzed to be 53
t
- i-consistent with the analysis for VANTAGE SH fuel. Figures 5.110 and 5.111 show the results. l The minimum DNBR remains above the limit. The conclusions of the FSAR remain valid. '
i loss of External Electrical Imad and/or Turbine Trin (FSAR Section 15.2.7) This event is analyzed to show that the DNB design basis is met and that primary and secondary side system pressures do not exceed 110% of design values. Four cases are analyzed which are: 8 BOL with pressurizer pressure control BOL without pressurizer pressure control EOL with pressurizer pressure control EOL without pressurizer pressme control The increased rod insertion time to the dashpot will not result in system pressures exceeding 110% of design values. Pressure transients from the current analysis of record were evaluated by extrapolation assuming the reactor trip was delayed 0.5 seconds. In all cases there wu ample margin to account for the slight expected pressure rise due to the slower drop times., ne increased rod drop time to the dashpot will not result in DNBR below the design limit. DNBR for the BOL C without pressure control and both EOL cases rises continuously throughout the transients. Therefore, the increased insertion time will have no effect on the minimum DNBR for these cases. DNBR during the BOL with pressure control case initially rises and then decreases to about the s initial value at the time of reactor trip for this case and the increased rod drop time will not result in a DNBR below the design basis. loss of Normal Feedwater (FSAR Section 15.22) This event is analyzed to show that adequate heat removal capability exists via the Auxiliary Feedwater System to remove core decay heat, stored energy and RCS pump heat following reactor trip. His is demonstrated by ensuring that the RCS heatup is turned around prior to the time when coolant expansion causes the pressurizer to become filled with water. The loss of feedwater i transient is a slow long term heatup event and is not sensitive to the rate at which control rods l are inserted following a reactor trip. The results of the current analysis of record and conclusions of the FSAR remain valid. l l-L 5-4 1
I i
?
e i Lnas of Ohite Power to the Station Auriliaries (FSAR Section 15.2.9) ! nis event is analyzed to show that' adequate heat removal capability exists via natural circulation flow as aided by the Auxiliary Feedwater System to remove core decay heat and stored energy following reactor trip. His is demonstrated by ensuring that the RCS heatup is turned around , prior to the time when coolar.t expansion causes the pressurizer to become filled with water. De RCS volumetric expansion is not affected since the total RCS flow and vessel outlet temperature remain the same. His transient is a slow long term heatup event and this aspect of this transient , is not sensitive to the rate at which the rods are inserted during a reactor trip. With respect to DNB criterion, this event is bounded by the Complete loss of Forced Reactor Coolant Flow analysis which was reanalyzed and shown to be acceptable. The conclusions of the current analysis 4 of record and the FSAR remain valid.
.ts Excessive Heat Removal due to Feedwater System Malfunctions (FSARbetion 15.2.10) '
His event is analyzed to show that the DNB design basis is met. De reactor is tripped on a turbine trip signal and minimum DNBR occurs shortly afterward. A conservative evaluation of the effects of the 0.5 second increase in control rod insertion time was done by extrapolating the ( transient DNBR results assuming that reactor trip was delayed by 0.5 seconds. He extrapolation showed that ample margin to the DNB limit still exists with a 0.5 second delay, ne conclusions of the F5AR remain valid. Excessive load Increase Incident (FSAR Section 15.2.11) This event is analyzed to show that the DNB design basis is met following a step load increase ' from rated power. Cases are analyzed at BOL and EOL conditions with and without rod control. In all cases analyzed, the reactor stabilized without a reactor trip. Therefore, the increased control l rod insertion time will have no effect on this event. De conclusions of the FSAR remain valid, i Accidental Deoressurization of the Reactor Coolant Svstem (FSAR Section 15.2.12) he limiting criteria for this event is the DNB design basis. This transient is terminated by a l reactor trip on Overtemperature AT. Minimum DNBR occurs immediately following reactor trip. ! A conservative evaluation of the effect of the 0.5 second increase in control rod drop time was done by extrapo!ating the transient DNBR results assuming reactor trip was delayed by 0.5 seconds. The extrapolations showed that abundant margin to the DNB limit still exists with a 0.5 second increase in rod insertion time. The conclusions of the FSAR remain valid. 55
I
~.'
he%tal Deoressurization of the Main Steam System (FSAR 5ection 15.2.13) and Maior Runture
. of a Main Steamline (FSAR Section 15.4.2.1) he inadvertent opening of a steam generator relief or safety vahe case is analyzed to show that the DNB design basis is met. De steam system piping failure cases are analyzed to show that the core remains intact and in place and that the radiation doses do not exceed the guidelines of 10CFR100. His is demonstrated by showing that the DNB design basis is met, even though DNB and possible clad perforation are not necessarily unacceptable for these cases.
he limiting steamline break cases are analyzed from hot shutdown initial conditions. De transient is started assuming the reactor is tripped and the core is at the minimum design shutdown margin. Derefore the 0.5 second increase in rod insertion time will have no effect on the results of this analysis. The conclusions of the FSAR remain valid. ., Snurious Oneration of the Safety Iniection System at Power (FSAR Section 15.2.14) e The spurious operation of the safety injection system produces a negative reactivity transient causing a reduction in core power. De power reduction causes a decrease in reactor coolant average temperature and consequent coolant shrinkage. Pressurizer pressure and level decrease until the reactor is tripped. During the transient the DNB ratio never decreases below the initial value, therefore the 0.5 second ir. crease in control rod insertion time will have no effect on the minimum DNBR. De conclusions of the FSAR remain valid. Sinele Rod Cluster Control Assembly Withdrawal at Full Power (FSAR Section 15.3.61 With the reactor in automatic or manual control, an upper bound of the number of fuel rods experiencing DNBR below the safety analysis limit is 5% of the total fuel rods in the core. Because the analyses takes no credit for the control rods entering the core at the time of trip, a
- 0.5 second increase in the RCCA drop time would have no effect on the results. The conclusions of the FSAR remain valid.
Maior Runture of a Main Feedwater Pine (FSAR Section 15.4.2.2)
'his event is analyzed to show that adequate heat removal capability exists using the auxiliary feedwater system to remove core decay heat, stored energy and RCS pump heat following reactor trip. This is demonstrated by ensuring that the RCS heatup is turned around prior to the time j
56
f i i i at which the hotlegs would become saturated and the peak primary and secondary pressures would exceed 110% of design values. The feedline break is a long term heatup event and is not sensitive to the rate at which the control rods are inserted following a reactor trip. De heatup transient continues for many minutes following the reactor trip. De 0.5 second increases in control rod insertion time will result in an insignificant increase in the integrated heat produced by the core during the transient. No i significant increase in holleg temperature or system pressures would occur due to the inecease in , control rod insertion time. ne conclusion of the current record of analysis and the FSAR remain valid. . Sincle Reactor Coolant Pumn Locked Rotor (FSAR Section 15.4.41 The accident is postulated as an instantaneous seizure of a reactor coolant pump rotor. Flow i through the affected coolant pump is rapidly reduced leading to an initiation of a, reactor trip on a low flow signal. If the reactor is not tripped promptly clad temperature may exceed the limit value of 27000F and RCS pressure may increase above that which would cause stresses to exceed the faulted condition stress limits. his transient can be very sensitive to the RCCA drop time. The locked rotor transient was reanalyzed with a RCCA drop time of 2.7 seconds. De results of thianalysis are shown in Figures 5.112 through 5.114 and Table 5.12. Since the. peak RCS pressure reached during the transient is less than that which would cause
' stresses to exceed the faulted condition stress limits, the integrity of the primary coolant system t
'is not endangered.
t:t @ m w .x u .
, Since the peak clad surface temperature calculated for the hot spot remains less than 27000F and the amount of Zirconium water reaction is small, the core will remain in place and intact with no .
L consequential loss of core cooling capability.
- j. .
Runture of a Control Rod Drive Mechanism Housine (FSAR Section 15.4.61 The RCCA ejection transients are characterized by a rapid power burst. Due to the speed at which the power increases this transient can be sensitive to the RCCA drop time. he RCCA transients were reanalyzed with a RCCA drop time (time to dashpot) of 2.7 seconds. De limiting 57
--w.. , - , , _ , ,_ ,.....e
, , i F ,
]
, transients were reanalyzed with a RCCA drop time (time to dashpot) of 2.7 seconds. The limiting j criteria for this event (References 12 and 13) are l
c .
- 1) Average fuel pellet enthalpy at the hot spot limited to below 225 cal /gm for unirradiated i fuel and 200 cal /gm for irradiated fuel. j a
L
- 2) Fuel melting limited to less than the innermost 10 percent of the pellet at the hot spot.
(Melting is assumed to occur at 49000F for BOL conditions and 48000F for EOL conditions) , i ne results of the analysis and a summary of parameters used in the analysis are shown in Table 5.11. All cases analyzed meet the acceptance criteria. Mana/Enerry Release (Steamline Breaks) (FSAR Section 6.2) . The mass and energy releases inside containment following a steamline break are used in the s containment integrity analysis and are insensitive to the rate at which the control rods are inserted. , The 0.5 second increase in control rod insertion time would increase the integrated energy produced by the core by an insigniGcant amount. De total RCS flow rate will be the same and no significant change in the overall system response will occur. Therefore, the conclusions of the FSAR with respect to containment integrity are not affected. he mass and energy releases outside containment following a steamline break are used to assure
~
that environmental conditions used for instrument qualifications are mai'n tained. As with the mass and energy released inside containment, the 0.5 second increase in control rod insertion time would
-increase the integrated energy produced by the core by an insignificant amount. The total RCS Dow rate will be the same and no significant change in the overall system response will occur.
Swi=line Break with Coincident Rod Withdrawal at Power his event is analyzed to show that the DNB design basis is met. he transient minimum DNBR occurs immed!stely following the reactor trip. De limiting case was examined with the new RCCA
! drop time of 2.7 seconds, and results were compared with those obtained from the presious RCCA drop time of 2.2 seconds, ne decrease in minimum DNBR was negligible and ample margin to l
( the DNB limits exists for the increased RCCA drop time of 2.7 seconds. 5-8 l'
3 4 1 i TABLE 5.11 ; C' Rod Custer Control Assembly Ejection Accident Results Time in Life BOL BOL EOL EOL i Power Imel 102' O 102 0 i Ejected Rod Worth 0.20 0.75 0.21 0.97
' (%AK)
Delayed Neutron Fraction 0.55 0.55 0.44 0.45 ; I Feedback Reactivity 1.3 2.4 1.6 3.63 . Weighting . Trip Reactivity (%AK) - 4.0 2.0 4.0 2.0 Fo before Rod Ejection 2.52 -- 2.52 -- ! Fo after Rod Ejection 7.11 14.05 7.88 2'6.0 Number of Operational Pumps 4 2 4 4 Max. Fuel Average 4097 3156 4001 3744 C Temperature (OF) Mx hil Center 4971 3610 4871 4391 Teq csycire (DF) hr Tuel Stored Energy 180 132 174 161
. (cal /gm) .i i
1 59 ;
p p
- l. (; TABLE 5.12 Summary of Results for Locked Rotor Transients Maximum Reactor 2603 Coolant System Pressure (psia)
Maximum Clad Temperature (OF) 2026 Core Hot Spot Amount of Zr.H 2O at Core 0.70 Hot Spot (% by Weight) 9 9 5 10
f FIGURE 5.1 1 j i. C Startup from Suberitical ! t ! Nuclear Power versus Time f f
- f. ,
i t.- I? i< I 10 1 n I h- , 1- o U !
'fw 30 i 0 .
I at C 3d : .
.g 10*l-- , .i g 1
e 10-2 0 2 4 6' S 10 12 14 16 18 20 22 21 le e 50 TIME ISEC1 I e*= 5-11 N 9 2
pw- - ., p MGURE5.12 !
'(.- l Startup from Suberitical R Heat Flux versus Time I-e a P
i i
- 1. i 9--
r
.B-- !
^ '
7 -
,'h w -
o g.6 w.5--
-5 -
g.4
.8 -
.2
.1 -- <
0. 0 2 4 t, 8 10 12 14 10 18 20 21 24 i t, if 30 TIME (SECS L f e 4 5-12
W
-y I
i RGURE 5.13 j i Uncontrolled Rod Withdrawal from a Suberitical Condition , i Fuel and Clad Temperature versus Time ' l 2000. !
, g 1600; ,
I i 1600. / 2 sk i f'1400. ; w. [ 1800.' ; W f- .
; c 1000. '
.3
~$.
- 6 800.
600. ) ; p 400. , D. 2. 4 6. 9. 10. 12. 14 16. 18. 20. 22. Ed. 26. 28. 30. Tirit t$tti ! r 900
?
b o : 750 < b g 700< < w I W m' 650 , g:. 600 u l' 550
$00 --
. . , 0 2 4 6 8 10 12 14 16 IB 20 22 24 E6 2B 30 i TIME ISECl l-5-13 l . .
i { k FIGURE 5.1-4 Partial Ims of Forced Reactor Coolant Flow Core and Loop Flow versus Time s ll. o 8.e
- 3. 8 '
- l. ,
t.' v'
= *%
l .e
' ' - ~ ~ _ Core Flow b
e-
~~~~.
e6 ' f
.g.
4
- Loop Flow 6
.t 8.
0 8 4 8 4 6 6 ? e 9 at flMt altte 1 4 5-14 1
y? . ,
' );
i fi f.1 i, .,'
~ .
f 4 j FIGURE 5.15
'5 i
Partial Loss of Forced Reactor Coolant Flow Heat Flux versus Time l 1 i l
, 1 i
s i 1.4 8 1.2 I ,. . b' y- ' t
)u * .
* .a.
W t
.t e.
* * * * . s . , . . i, unt isce 5-15 S
=
! . l.-
i
'$ l I
MGURE 5.16 0 f\ ic Partial las of Forced Reactor Coolant Row DNBR versus Time i
'p.
l 1 I a l j.. ,
- ?!, +
l t.s i l' t.e< 4 l 1.8 j J 1 a
. 8.
I
~
i l.D 86' 34 , F4 8.8 8 3 8 3 4 5 6 ? e g 3e . Tiet eggce e 5-16
- u. T .(
- k. 4
3 1 l I
, , 1 i
MGURE 5.17 i l j '. i Complete loss of Forced Reactor Coolant Mow l Core and Loop Flow versus Time
]
l c. i l 1 4 L l l i l l 1 1.4 1 6 1' i i' 't.t r
- i. .
D C 6
.J g.
m 4' g E
& .t. a 4
,. , 8. .
0 8 8 8 4 5 6 7 8 9 30 1888 Istti 5 5-17 4 ) ,
< g -.
q l 1 I l FIGURE 5.14 < [_ , . Complete Ims of Forced Reactor Coolant Flow Heat Flux venus Time . i I l
)
\l 'i
- l k-I, ~
1 (
)
d I (' i I j,
&.e I
lee < }' ~ y, 5 a-e
'6
.h t
E 4 $ t Y L e$
.4.
8 I & 4 4 5 6 7 8 9 le
, i flat IStti l
?
e L e V
+
5-10 .
.'4 i
vr::q
. . ,x
- g ,
p - r o, :;09~ w * 'v';.
, .y j
? l
~'y '
'V
. MGURE 5.19 '
- t. .
f, Complete Loss of Forced Reactor Coolant Fiow ! T DNBR versus Time . s 4 97- -
- E.
1
. ),: L s.
i , i 3 s ., n> ;,;l
+
f.6
+
.t.4 6 4
tst
, .b.. . ' . , . .to
(
%- '. f 7p 3*6 s,'
't.4 1.
- 3 2 3 4 't ? g
, 6 e go s
s TIME ISECD o-
..I' i
~
a
- 3 ._
e 9 9 S-19
.u 4
.~.2_.m_ -
,Y
- n. .
{ r.
~ '
1 { , '_\
!l- s
>s
) g g e E. d',k ' :
N FIGURE 5.1 10 - [ b >- , Startup of an Inactive 1. cop W
,(
Heat Flux versus Time t' 4 llI, - l5', - (:-: t i. lp &.".' ? t,
, _ F t.
.it . 4
!. 9, '.
T
,;L ' S*2 l
-(
E' .i. i L. Q. 4 u .s. L' u 3 I, .* _6
'> . F q.
.. W. ;, e 4-
.\
N- -..- c. O' 2 4
. 6_ t l' 42 14 la il 26 fIME titti P
T
~
l
^ . o
.
- 32 l'
5-20 1
- l 1
1 l 4 1
, , 1 r
>. 6,, 3 e
. ... '![: .
jF E g
~'
, .y MGURE 5.1 11 ~. t p- _ .p Startup of an Inactive Loop DNBR versus Time . j;; , t (
)
M-s l s, at Thimble Cell
- t.8
+
t.6 t.4 3 ,
. t
-j:i 2.2*
h a r
$ 4 ,: -
e f P 4.8 t i,. - .
.t t
6 3.4 =. 5 I.t
- 9
g 3 4 4 e t
'ftnt 'IEC8 c
i 5-21
.F-
, m 4f,
, a:- -
F'. (. .
,i 3.
y a s (' . FIGURE 5.1 12 Single Reactor Coolant Pump Locked Rotor Core Flow versus Time: t-I i' h l. t
. I ~. 4 I'8'
,h'
*"'.3.
f-
.j .. .
c C.
.a l
.6 j M.
m-Y
~
, s 4
i
'd-. .2 l
\
l
- 0. .j
., , b 8 8 3 4 6 6- ? e 9 to j fint IStC)
I l:. I: 1 1 i l 1 '.- l 5-22 L I b
s
]
l t i.; . ( - FIGURE 5.1 13 Single Reactor Coolant Pump Locked Rotor
' Nuclear Power and Heat Flux versus Time .
I l i.. t l
._St
^
I 1. b 3
= .
$ .6 g ,
w l
.t
- 8.
.. . t . . . . . . 9 ..
Timt- ($tts 8..
- 1. 2
- I ..
8 y .. 2 g .. , m y ..
.t 8.
0 3 3 3 , I 6 7 8 9 10 T!nt f$tCe 5-23
7r 39' y >, g' / ~ -
. e k, . *
>l . k :o y y,
'j s
D.. . ' FIGURE 5.114 ' c Single Reactor Coolant Pump Locked Rotor . Clad Temperature versus Time 1 a ': 2
,1 l
l 6 + t ' i e
-2200. s 2000. ,
'.'u. 3000. .
. d. -
. 3600.
W-gg 1400. C .-
.300 1000.
800. I
< 600.
8' 3' 8. 8. 4 5. 6. 7. 8. 9. 80.
? ! rit 8 Sits r
a i S-24
x ; 5.2 LOCA Accidents
*Ihis section summarizes the evaluations performed to assess the effect of VANTAGE SH fuel on
~
the Sequoyah Units licensing bases LOCA analyses. As noted in Reference 1, the majority of the VANTAGE SH fuel features have no adverse effect on the licensing bases LOCA analyses due to the mechanical and hydraulic similarity to 17x17. STD fuel. Reference 1 documents that transitioning from 17x17 STD fuel to 17x17 VANTAGE SH fuel without intermediate flow mixers l (IFMs) results in no transition core Peak Cladding Temperature (PCT) penalty. The only item .i which can potentially affect the LOCA analyses is the increase in RCCA rod drop time. Larne Break LOCA (FSAR Section 15.4.1) The Large Break LOCA analysis for the Sequoyah Units has been executed using the Westinghouse 1981 Evaluation Model with BASH for 17x17 standard fuel. It resulted in a calculated PCT of 20010F for the limiting Cd=0.6 DECLG break. This analysis considered the upper head injection (UHI) system to be removed from senice. ( An evaluation has been performed to consider the effects on the analysis of the VANTAGE SH fuel. The large break LOCA evaluation model does not take credit for the negative reactivity introduced by the control rods. Instead, the reactor is brought to a subcritical condition by the presence of voids in the core caused by the rapid depressurization of the RCS. Since credit is not taken for the negative reactivity introduced by the control rods, the increase in rod drop time will have no effect on the current FSAR large break analysis. Furthermore, sensitivity studies have demonstrated that VANTAGE SH fuel is less limiting than 17x17 STD fuel in analyses performed using the 1981 Evaluation Model with BASH. Based on the discussion given above, the use of VANTAGE SH will not result in an increase in the peak clad temperature for the Sequoyah Units. Therefore, these changes are acceptable and the resulting peak clad temperature remains within the regulatory limits. l !- Small Break LOCA (FSAR Section 15.3.11 The small break LOCA licensing analysis for the Sequoyah Units for UHI removal has predicted a peak clad temperature of 2105.50F using the NOTRUMP Westinghouse Small Break Evaluation Model. The only VANTAGE SH feature which affects the Small Break LOCA analysis 5 25
1- f _q is the increase in rod drop time. He Westinghouse small break model assumes that reactor core 1 is brought to a suberitical. condition by the negative reactivity of the control rods. He increase in the rod drop time to a maximum value of 2.7 seconds has been modeled in the UHI removal . ; small break LOCA analysis. There is no increase in clad temperature due to the introduction of VANTAGE SH fuel. De revised UHI removal licensing basis for the Sequoyah Units would therefore be applicable independent of the use of VANTAGE SH fuel. Steam Generator Tube Ruoture (FSAR Section 15.4)
. The steam generator tube rupture (SGTR) accident is analyzed to ensure that offsite doses remain
, below the limits defined in 10CFR100. He primary thermal hydraulic factors affecting this conclusion are: the extent of fuel failure that occurs during the event (or whether DNB occurs),
the primary to secondary flow through the ruptured tube, and the mass and energy releases to the atmosphere from the steam generator with the ruptured tube. The amount of fuel failure , assumed for the SGTR analysis in the Sequoyah FSAR is 1%, which is assumed to be independent of the fuel type or the transient conditions. The primary to secondary break flow and the mass ,. released to the atmosphere from the ruptured steam generator are dependent upon the RCS and secondary system operating parameters. The change from 17x17 STD fuel to VANTAGE SH fuel
' does not affect these parameters and therefore has no effect on the SGTR analysis. -
Blowdown Reactor Vessel and Iron Forces (FSAR Section 3.9.3,5) l The major factors in determining the resulting forces from a postulated LOCA on the vessel and 1 the internals are the reactor coolant system primary fluid temperature and pressure. Since i VANTAGE SH does not change the primary side design temperatures and pressures which are : modeled in the forces analysis, there will be no effect on the LOCA hydraulic forces. L Post LOCA lonn Term Core Cooline. Boron Evaluation (FSAR Section 15.4.11 ne Westinghouse licensing position for satisfying the requirements of 10CFR Part 50 Section i l l 50.46 Paragraph (b) Item (5) *long-Term Cooling' is defined in WCAP-8339. The Westinghouse ; commitment is that the reactor will remain shutdown by borated ECCS water residing in the RCS and sump after a LOCA. Since credit for the control rods is not taken for a large break LOCA. the borated ECCS water provided by the accumulators and the RWST must have a concentration F that, when mixed with other sources of borated and non-borated water, will result in the reactor l 546
; core remaining subcritical assuming all control rods out. This will be demonstrated for each reload
' design with VANTAGE SH fuel on a cycle specific basis.
Since the use of VANTAGE SH fuel will not affect the sources of borated and non borated water L assumed in the long term cooling calculation, it is concluded that there would be no change to the long term cooling capability of the ECCS system. As noted above, this licensing commitment is checked by Westinghouse on a cycle by cycle basis ensuring compliance with this requirement independent of this safety evaluation. Hot 12e Switchover to Prevent Potential Boron Precioitation (FSAR Sections 15.4.1 and 6.31 Post LOCA hot leg recirculation switchover time is determined for inclusion in emergency-procedures to ensure no boron precipitation in the reactor vessel following boiling in the core. This recirculation time is dependent on power level, and the RCS, RWST, and accumulator water volumes and boron concentrations. The VANTAGE SH fuel will have no effect on the assumptions for the RCS, RWST, and the accumulators in the hot leg switchover calculation. Thus, ' there is no effect on the post-LOCA hot leg switchover time. Rod Eiection Mass and Enerev Release for Dose Calculations (FSAR Section 15.5.7) A review of the Sequoyah Updated FSAR Chapter 15.5.7 shows no LOCA analysis of the primary side mass and energy release due to a hypothesized rupture of a control Rod Drive Mechanism
- (CRDM). Instead, simplifying assumptions were made to determine the release of radioactive material from the primary side as a result of a CRDM rupture. Therefore, the use of VANTAGE SH fuel at the Sequoyah Units will have an insignificant impact on the calculated consequences of a rod ejection accident and is acceptable.
i 1 l 5-27 l
e-3 5.3 Environmental Consequences of Accidents I This section summarizes the impact of the upgrade to VANTAGE SH fuel for the Sequoyah Umts on the radiological consequences of accidents, ne change to VANTAGE SH fuel affects the accident doses only insofar as the extent of fuel damage is increased and/or the coolant mass releases to the environment are increased.
' Included in this evaluation is consideration of extended fuel butaup of 48,000 f. 500 MWDMTU for the reload batch average discharge. This extended burnup has a peak fuel pin burnup of
< 60,000 MWDMTU.' De impact of extended fuel burnup on core source terms used in accident analyses is addressed in References 3 and 11. . Based on Reference 11, the extended fuel burnup would have no impact on the radiological consequences of any of the design basis accidents, except for the fuel handling accident, because the doses for these accidents are due'to the release of short lived isotopes of krypton, xenon, and iodine which, because of their short half lives, do not increase with burnup. There is an increase in Kr-85 inventory associated with extended burnup but Kr 85 does not significantly impact radiological consequences.
The reactor coolant system source terms provided in FSAR Section 11.1 are not significantly . affected by the implementation of extended fuel burnup. While a few long lived isotopes would increase appreciably, most isotopes would remain virtually unchanged and others would be reduced. The operation of the reactor would remain limited by Technical Specification 3.4.8 which requires that the specific activity of the primary coolant be limited to less than or equal to 100/E microcuries per gram. l Environmental Conseauences of a Postulated loss of A.C. Power to the Plant Auxiliaries (FSAR Section 15.5.1) The VANTAGE SH fuel features do not result in any fuel damage associated with this accident L nor is there an increase in the mass releases from the secondary coolant system. As discussed L above, extended fuel burnup does not increase the source terms associated with this accident. He
- conclusions of the FSAR remain valid.
l 6 5-28
J / 1 Environmental Conseauences of a Postulated Waste Gas Decav Tank Ruoture AL, (FSAR Section: 15.5.2) i ne VANTAGE 5H fuel features do not affect this accident since it involves an auxiliary system that is separate from the operation of the reactor. As discussed above, extended fuel burnup does s i I1 not increase the source terms associated with this accident. The conclusions of the FSAR remain ! 1 valid. i l Environmental Conseauences of a Postulated Ines of Coolant Accident (FSAR Ser; tion 15.5.3) ne VANTAGE 5H fuel features do not result in any increase in the fuel damage associated with determining the radiological consequences of this accident since the level of core damage assumed is based on the guidance of Regulatory Guide 1.4. If a mechanistic determination of the fuel damage were to be used, with or without the implementation of the VANTAGE SH fuel features, there would be a reduction in the level of fuel damage. Mass releases from the secondary coolant system are not considered for this accident. As discussed above, extended fuel burnup does not increase the source terms associated with this accident. The conclusions of the FSAR remain valid. Environmental Conseauences of a Postulated Steam Line Break (FSAR Section 15.5.4) he VANTAGE SH fuel features do not result in any fuel damage associated with this accident nor is there an increase in the mass releases from the secondary coolant system. As discussed above, extended fuel burnup does not increase the source terms associated with this accident. The conclusions of the FSAR remain valid.
~
Environmental Consecuences of a Postulated Steam Generator Tube Ruoture (FSAR Section 15.5.51 + The VANTAGE 5H fuel features do not result in any fuel damage associated with this accident nor is there an increase in the mass releases from the secondary coolant system. As discussed above, extended fuel burnup does not increase the source terms associated with this accident. The conclusions of the FSAR remain valid. Environmental Conseauences of a Postulated Fuel Handline Accident (FSAR Section 15.5.61 The VANTAGE SH fuel features do not affect this accident. However, based on Reference 11, the implementation of extended fuel burnup would result in an increase in the gap fraction assumed 5-29
1
. for I 131 from the ten percent value specified in Regulatory Guide 1.25 to twelve percent. Since 7\ I I 131 is the primary contributor to the thyroid dose, this would be expected to increase the thyroid dose by twenty percent. ,
While it is expected that analysis of the I 131 gap fraction for the fuel management scheme specific i
. to the Sequoya t fuel would show that the peak rod gap fraction does not exceed ten percent, the ' ;
fuel handling accident doses have been reanalyzed utilizing the increased gap fraction value. The - analysis bases have also been revised to counteract the impact of the increased I 131 gap fraction. There are two changes. One involves the use of accident meteorology that is based on a larger data base than was used previously (and thus more accurately reflects the Sequoyah site). The f other involves the use of Regulatory Guide 1.52 guidance for charcoal filter efficiency credit in place of the guidance provided in Regulatory Guide 1.25. The results of the reanalysis show that for the fuel handling accident occurring in the Auxiliary Building the thyroid doses reported in the
- FSAR of 71 rem' at the Site Boundary (SB) and 8.3 rem for the Low Population Zone (LPZ)- ,
remain bounding. For_ the fuel handling accident located inside the containment the reanalysis e shows that the SB thyroid dose of 118 rem reported in the FSAR remains bounding but that the LPZ thyroid dose of 13.8 rem is exceeded. The revised LPZ thyroid dose is less than 20 rem. The conclusions.of the FSAR remain valid (i.e., that the doses for the fuel handling accident remain below the limits of 10 CFR 100). Although the LPZ dose has increased somewhat for the case in which the fuel handling accident occurs inside the containment, that dose is less than the acceptance criterion of the present Standard Review Plan and is less than the reported doses for E the SB. Environmental Consecuences of a Postulated Rod Eiection Accident (FSAR Section 15.5.7) The VANTAGE SH fuel features do not result in any increase in the fuel damage associated with this accident nor is there an increase in the mass releases from the secondary coolant system. As discussed above, extended fuel burnup does not increase the source terms associated with this . accident. The conclusions of the FSAR remain valid. l 1 5 30
f( W < , T l
6.0 REFERENCES
- 1. ~ Davidson, S. L ed. et al., " VANTAGE SH Fuel Assembly," WCAP-10444 P A, Addendum ,
o ~2A, April 1988.
- 2. Davidson, S. L ed. et al., " VANTAGE 5 Fuel Assembly Reference Core Report," WCAP '
- 10444 P A, September 1985.
o
- 3. Davidson, S. L ed. et al., " Extended Burnup Evaluation of Westinghouse Fuel," WCAP L i 10125 P-A, December 1985.
- 4. Davidson, S. L ed. et al., " Westinghouse Reload Safety Evaluation Methodology,"'WCAP-9273-A, July 1985.
i
- 5. Skaritka, J., " Operational Experience with Westinghouse Cores," (through December 31, 1988), WCAP-8183, Revision 17, August 1989. ;
C 6.' Weiner, R. A., et al., " Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations," WCAP 10851 P-A, August 1988.
- 7. F. E. Motley, K. W. Hill, F. F. Cadek, and J. Shefcheck, "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," WCAP-8762-P A, July 1984.
8.- Skaritka, J., ed., " Fuel Rod Bow Evaluation," WCAP-8691, Revision 1 (P), July 1979.
- 9. " Partial Response to Request Number 1 for Additional Information on WCAP-8691, Revision 1," letter, E. P. Rahe, Jr. (Westinghouse) to J. R. Miller (NRC), NS EPR 2515, dated October 9,1981; " Remaining Response to Request Number 1 for Additional Information on WCAP-8691, Revision 1," letter, E. P. Rahe, Jr. (Westinghouse) to J. R.
Miller (NRC), NS-EPR-2572, March 16,1982. 6-1
y A3
- 9 J
. 10.- letter from C Berlinger (NRC) to E. P. Rahe, Jr. (Westinghouse), " Request for Reduction C in Fuel Assembly Burnup. Limit for Calculation of Maximum Rod Bow Penalty," June 18, 1986.
1 4ov., 11. D. A. Baker, W. J. Bailey, C E. Beyer, F. C Bold, and J. J. Tawil, " Assessment of the Use of Extended Burnup Fuel in Light Water Reactors," NUREG/CR 5009 (PNL-6258), February 1988.
- 12. Risher, D.' H., Jr., "An Evaluation of the Rod Ejection Accident of Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1A.
- 13. NS NRC-89 3466, "Use of 2700 F PCT Acceptance Limit in non LOCA Accidents," W. J.
Johnson (Westinghouse) to Mr. Robert C Jones (NRC), October 23,1989.
\
l 1 1 b. l l l C l 1 6-2
v,, 1 f 51 "
.'l. f ' y 5 4
f i'
,5.,, ,
sty , in * ,, , () kt ,
\Y;i F
I;' ' r
- f. '
p: u- . APPENDIX A qJ - , Summary of Technical Specification Changes - Y l r.' ,. ,i L f *
?. , i n s .
- 12. '
g* !E
- s '
1 /
.5'f( d i
+>
th
.. p .'
e Y. l [ .- l l' b l I h
,,r _ _ _ , _ - _- . _ _ ,__ - , _ , _ _
1 y , 3
SUMMARY
OF TECHNICAL SPECIFICATION CHANGES , FOR SEQUOYAH UNIT 1 4 Page' Section Description . .
. B 2-1 ' 2.1.1 Basis - Change W-3 correlation This change reflects B23 to WRB-1 correlation and the DNB correlation B 2-5. added safety analysis .used in analyses. ,
B 3/4 2 5 3/4.2.5 Basis DNBR limit. - 3/4119 3.1.3.4 Revised rod drop time This change is a result-to less than or equal to - of changes in the fuel due y 2.7 seconds.' to the VANTAGE SH fuel design. The effect of this j i increase on the safety analysis 1 has been considered. 1 1
-3/4210 ' 3/4.2.3 . FAH and Rod Bow Use of the new rod bow penalty 3/4211 Penalty (delete Rod Bow' methodology reduces the rod bow. l
- 3/4 2-12 Penalty (RBP) as a function penalty. The reduced penalty is i 3/4 2-13 of burnup in FAH equation accounted for in the analysis by .!
.3/4214 _ _
and delete figure 3.2-3), using available DNBR margin. l B 3/4 2-13/4.2.2 and ' The new methodolgy is defined 3/4 2-2 3/4.2.3 Basis . in the references below . 1 B 3/4 2-4 q C 3/4218 3/4.2.5 DNB parameter This change is a result of 3/4 2-19 the revision of the B 3/4 2-5 3/4.2.5 Basis FAH Tech Specs with ' respect to flow. Skaritka, J., (Ed.) " Fuel Rod Bow Evaluation," WCAP-8691, Revision 1 (Prop), July 1979. Partial Response to Request Number 1 for Additional Information on WCAP-8691, Revision 1" letter, EP. Rahe, Jr. (Westinghouse) to J.R. Miller (NRC), NS-EPR 2515, dated October 9, 1981; " Remaining Response to Request Number 1 for Additional Information on WCAP-8691, Revision 1" letter, E.P. Rahe, Jr. (Westinghouse) to J.R. Miller (NRC), NS EPR 2572, dated March 16,1982. 12tter C. Berlinger (NRC) to E.P. Rahe, Jr. (Westinghouse), " Request for Reduction in Fuel Assembly Burnup Limit for Calculation of Maximum Rod Bow Penalty," June 18, 1985. 1
i
.r x, .,
. q.
e-
,.. .. i n ,?}',
t pt y a 4
-.g- ,
3 Recommended Modifications to r-the Technical Specifications for Sequoyah Unit 1 . ff i i 7
- 'd1 ' .
}
1 5 s 1 9 g
\
9 4 I \ i [ .. I t. t
k , " e . i 2,1 $4FITV LIM}TS-
,. BASES >
WRB-1 cortsjaton ord W .5 C4teddein b-t.od%m outsde ifna rwe. of WKB-1 L " The restrictions of this' safety limit pmv'ent overtisating of the feel and possible cladding perforation which would result in the release of fission k p, , y'
- producta-to the reacter coolant. Overheating of the fuel cladding is prevented p-ty restricting fuel oppration to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is- f. '
a slightly above the coolant saturation temperatum. y Operatian above the upper boundary of the nucleate boiling regias could E,,
# result in excessive elsedin
. 'from nucleate boiling (DNB)gand temperatures the resultant because of the enset sharp reduction of departure in host transfer coefficient. DNB is not a directly measurable parameter during operation and o
therefore THERMA. POWER and Anactor Coolant Temp ute and prestiure have.been related to DNB m rough EnT g N. . ' 2"d developedtopredicttheDNB@fluxandthefocati W DNB IC5Prvtgh&ggjhn ) o DN N ax1a ly uniform o J and non-uniform heat flux distributions. The local DNB heat flux ratio DNBR,
; defined as the ratio of the heat flux that would cause ONB at a particular
-core location to the local heat flux, is indicative of the margin to DNB. '
]'gt oper The minimum va of the DNBR-duri aatty state ope ion, normal
< nel transients, anticipated ra ts is liette 1.30. This (1 value ce that 0 5 wi ponds'to a 95 ent probability a t occur and is esen as an appropr operatine cond lens.
g5 percent con margin to DN nee level r all ; The curves'of Figure 2.1-1 show the loci of points of THERN R-p0WER, Reactor Coolant Systen isure and average temperature for.which the alls l sinimum DNBR is no less tha[nis , or theaquel averinetoenthalev the enthalpy at the vesse1 ~ edt of saturated liouid. .! the safety analysis DNBR limiti Thecurvesarebasedonanenthalpyhotchannelfactor,Fh,of1.55and - a reference cosine with a peak of 1.55 for axial power shape. - An allowance is included for an increase in F" g at reduced power based on the empression: E118l F ," = 1.55 [1+ 0.3 (1-P)) allsg where P is the fraction of RATED THERME POWER
.-(
SEQUDYAN - LatIT 1 8 2-1 Amendment No. 19, 114 May 5. 1989
?
i 9 INSERT 1 The DNB design basis is as follows: there must-be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 or W-3 correlation in ' this application). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit (1.17 for the WRB-1 correlation). l. L
)
l 1 1
. . . - - . - _ . . - - . ---w -. --_ - __ - -_ _ _
c
+. )
1 l
, SAFETY LIMITS -
C 11 J SASES-_ i
- I Manual Reactor Trio instrumentation channels and provides manual ~ reactor tri PowerRance.NeutronF1dx The Power Range, Neutron Flux channel high setpoint provides reactor core L
protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. L redundant n Llow power. protection in the power range for a powe r excursion beginning Thefrom low set pointl t L The trip associated with the low setpoint may be manually bypassed Li when P-10 is active (two of the four power range channels indicate a power ! level of above approximately 10 percent of RATED THERMAL POWER) and is auto- l8, L
'matica11y reinstated when P-10 becomes inactive (three of the four channels indicate a power level below approximately 9 percent of RATED THERMAL POWER).
L power Rance. Neutron Flux. Nich Rates (. The Power Range Positive Rate trip provides protection against rapid flux L increases which are characteristic of rod ejection events from any power level. ! Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for ead a4*etina fram nartial ==er. it>e
.The Power Range Negative Ra WNovidesprotectiontoensurethatsne. safutv minimum DNBR is maintained above for control rod drop accidents. At high L
i when in conjunction with nuclear power being maintained eq power by action of the automatic rod control system, could cause an unconserva-tive local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor for all single or multiple dropped rods. Inters.ediate and Source Rance. Nuclear Flux L z The Intermediate and Source Range, Nuclear Flux trips provide reactor
- < core protection during reactor startu).
tion to the low setpoint trip of the lower Range, Neutron Flux The channels.The Source Range Channels will initiate a reactor trip at about 100 counts per second unless manually blocked when P-6 becomes active. The Intermediate 1 1 SEQUOYAH - UNIT 1 8 2-3 Revised 08/18/87
- --- ----,-r - ,m ---- -
SAFETY LIMITS-BASES t analyses; however, its functional capability at the specified trip setting is
-required by this specification to enhance the overall reliability of the Reactor Protection System.
Pressurizer Pressure
- The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High Pressure '
trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of.a loss of reactor coolant pressure. , a Pressurizer Water Level L The Pressurizer High Water Level trip ensures protection against Reactor
- Coolant System overpressurization by limiting the water level to a volume
? sufficient to retain a staan bubble and prevent water relief through the L .; pressurizer safety valves. No credit was taken for operation of this trip in j: the accident analyses; however its functional capabil(ty at the specified
.tripsettingitrequiredbythIsspecificationtoenhancetheoverallreliability of the Reactor Protection System.'
Loss of Flow-L The Loss of Flow trips provide core protection to prevent DNB in the 14 v event of a loss of_ one or more reactor coolant pumps. Above 11 percent.of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 89% of nominal full loop flow. Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 89% of nominal full loop flow. I latter trip will prevent the minimum value of the DNBR from going below L during normal operational transients and anticipated transients when 3 loops areinoperationandtheOvertemperatureDeltaTtripsetpointisadjustedto the value specified for all loops in operation. With the Overtemperature Delta T trip set point adjusted to the value specified for 3 loop operation, the P-8
. trip at 76% THERMAL POWER will prevent the minimum value of the DNBR from going below during normal operational transients and anticipated transients with 3 loopshin operation.
/
the safety analysis DNBR limit SEQUOYAH - W IT 1 B 2-5
9
, A s .
l.. . e , i REACTIVITY CONTROL' SYSTEMS 7 p -RODDROPT!g 2.7 LIMITING CONDITION FOR OPERATION
~
'3.1.3.4' The individual full, length (shutdown and control) op time from the fully withdrawn position shall be less than or equal to seconds from m1121 si beginning of. decay of stationary gripper coil voltage to dashpot entry with: l
- a. 'T,,, greater than or equal to 541'F, and
+ 4
' b. All reactor coolant pumps operating. -.
APPLICABILITY: MODES I and 2-ACTION: ,
- a. ~ With the' drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
s- b. l1 With the rod drop times within limits but determined with 3 reactor 3 coolant pumps operating, operation may proceed provided THERMAL '
- ' - POWER is restricted to less than or equal to 71% of RATED THERMAL POWER'
-. . )
, SURVEILLANCE REQUIREMENTS 4.1.3.4 :The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:
a. For all rods following each removal of the reactor vessel head, b. For specifically affected individual rods following any main- '
* ' tenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and
- c. At least once per 18 months.
~
# Fully withdrawn shall be the condition where shutdown and control banks are at a position inclu'ive.s within the interval of > 222 and < 231 steps withdrawn,
~ ~
1112 L SEQUOYAH - UNIT 1 3/4 1-19 Amendment No. 108 March 28, 1989
7 1 3/4.2.31MUCLEAR ENTHALPY HOT CHAWEL FACTOR- Fj J POWER DisTRrBurIOw units 3.2.3. Fd shall be Jimked by b
> Y 4. h th F M T M R 0 Mb ! "'
~
L!n!TINGCONDITIONFOROPERATION .- 3.2.3 The e ination of icated Reac Coolant Sys h (RCs) total 'I ow rate and .R2 shall intained wi intheregion{ofallowable h i ope tion shown o igure 3.2-3 r 4 loop ope tion: \
, x 1
1.49 [1.0 + 0 3 (1.0 - PN
+' Fgg 6l.55 09%3Q.0-V y,
R, s*. . [1 NRBP (Bu)] , THERMAL POWER , RATFD TNFRht&l pnWFD 18 *
~'~
. Fh = Measured lues of Fg obtained by sing the able incere dete rs to obta n a power d tributi'n o .
measured v lues of Fh hall be'use to calcula Rs e Figure 3. -3. include measurement neertainti of 3. for flow a 45forineremeasurenetofFka
- e. R8P (Bu = Rod Penalty a a function f region av age burnup s shown in igure 3.2-4, where a reg n
, is deft d as those semblies w h the same Ioading d $e (reloads) r enrichseht (first co .
, APPLICABILITY: MDDE 1 ACTION: .
th the c tion of RCS tal flow r and Rg , outside th regions o cceptable ration 'shown Figure 3. : p tiithin- urs: L 1. Eithe store the c ination of CS total f rate L and Rj , to within above limi or I ,{nggpf p, 2. educe ME R and red POWER to le than 50% o RATED D E l the Power Neutron ux - High tr setpoint less than or .ual to 555 RATED THE L POWE ithin the ~ _ t 4 hours.. K m December 23, 1982
, SEQUOYAH - UNIT 1 3/4 2-10 Amendment No. 19 -
~
E-l:'_ ..; E l INSERT 2. L With FNg exceeding its-limit:- l
. .a.- . Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High Trip i
setpoints tio 5 55% of: RATED THERMAL POWER within the next 4 hours,
- b. - Demonstrate thru in-core mapping that FgN is within its' limit L- . within 24-hours after exceeding the limit or reduce THERMAL POWER- !
l to less than 5% of RATED THERMAL POWER within the next 2 hours,. o and l? c.. Identify and correct the cause of the out of limit condition
-prior to increasing THERMAL POWER above the reduced' limit l required by a. or b. above; subsequent POWER OPERATION may.
proceedprovidedthat(Hisdemonstratedthroughin-core ! mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of i RATED THERMAL POWER prior to exceed'ing this THERMAL power and L- ' within 24-hours after attaining 95% of greater RATED THERMAL
~
POWER. \; L.
7 POWER DISTRIBUTION' LIMITS ON:' (Conti d) '
. - lb. Within 24 urs of in ially being' utside the ve limits, '
orify thro h incore f1 mapping RCS total ow rate c arison tha the combi ion of Rg , ' and RCS to 1 flow rate re resto to within above li ts, or_ reduc THERMA POWER to le than 5% o RATED THE POWER with the next hours. ~
- c. ntify'and erect the e se of the o -of-limit ndition pri to increas THERMAL R above t reduced RMAL .
POWER imit requi by ACTION tems a.2 a or b. above, subseque t POWER OPE TION say p eed provide that the '
. .. combinatio of Rj , R a indicated CS total f1 rate are ,
NII 2 nstrated, through inc e flux mapp .and RCS tal flow
, ra comparison, to be with the region of acceptab opera on shown,i Figure 3.2- prior to e eeding the \N ,
1 followi 'THERNAL R 1evels:- -
. A nomi 50% of RA ' THERMAL ER, 2.- A nominal of RATED ERMAL POWE and
' 3. hin 24 hour of attaini greater or equal 95%
N of TED THERMA POWER. SURVEILLAN REQUIREME S s s s s s s 4.'2.'.1 The pro isions of cification .0.4 are no applicable. s 4.2.3.2 The combi ion of ind ted RCS to 1 flow rat and R , R j 2 shall be termined to within regions o cceptable eration gure 3. -3: SEQUOYAH - UNIT 1 3/4 2-11
.ep 9
e'~=--
5
. : r- . g ,
. . J u D 5er. s .
- ; c l.
.SURVE!LLANCE REQUIREMENTS' L- . i l
4.2.'3.1 The provisions.of Specification 4.0.4 are not applicable. 1-4.2.3.2 . F"g shall be detemined to be within its 11 alt by using the novable incore detectors to obtain a power distribution sep:. , L a. Prior to operation above 755 of RATED THEREL PWER after each fuel loading, and ! b. At least once per 31 Effective Full Power Days.
- c. The measured Fh shall be increased by 45 for esasurement ,
uncertainty. * ' s o G I-
i I f ';
. POWER DISTRIBUTION LIMITS LLANCE REQUIREMENTS nued) a.- Pr to ration above 755 TED THERput fter * -' ~
- ~
t each f oading, and-
~
- b. At least once pe 1 Effective Full Power s. ..
. 4.2.3.3 The i sted RCS total rate shall be verift be within the region of accep le operation of re 3.2-3 at-least onc r 12 hours .
n the values of R j R ~, obtained per ification 4.2.3.2, a ssumed - ist.- - l 4.2.3.4 RCS. total flow rate i itors shall be s tad to a 1 CHANNEL CALIB at least once per nths. ! 5' The RCS flow shall be detemined asurement at- -1 least. per,18 months. , w -
.g O I C .
a-
. t L
-P e
E ( g .. l l
~
SEQUOYAH - UNIT 1 3/4 2-12
4 ' s' u
~
- p. j. .
'o e. .
'/g,-
t'
- - m ::: :: a m e s
y, g g_ * ' m I' t aM. _ f g y g W -
-i -miBliEERIII :
Ers=i r_a m:_ Q
- a iiii!!EBillilL gggE3!Q i L2)iii_!!='- -
g
,. _ _ __ _.= . ,
=n R
.gs:
. - s
~~'q* $,__,g' g...gnt g
m
. m._
__ a s et-
--= , ,.
_ i ..
=__
=m--
me=-es == me=e== ; . g) is" "m_ . - =m===; e._ l gn %, p- g q, ,l , w x .
-H g
~ + ,
O
-;lla -
e. p -
% %.g
- . IE. s= h.
c , G
- a g.
- - ~ ~
,,}>-
v a I
- <lI-
- g 'E g ga =.__
~
=
h l8.*5I. K ?" 1555i Hir - a g- g ggI 7
. g . y m :
5,sa e i '
- a '
e s s :- a ;. " .3 iredtpH RAVuaROld1V101SDW '
- q. -
- December.23, 1982 3/4 2-13 Amendment No.19 SEQUOYAH - UNIT 1 ,
w - , f -
\
, } ., j - d
]
x. f . 2 -
' t, .
j O l N.- . g , ,.
,,hnUJ":,,'i;, ,. .: , .
,i dillCh3"i%,l'i spht':0 6' . ,, 4 ,L',',,",
- WJL1Mt:ll:MilM@h + ,,
% g p,,1;" 164tn;'t. ,;.;N"t,", . - '
e e: "
- g";,! pi! ' .co;dt , , 'N .
' M6t'd'lhill MwW', 'ti 4 '
Un 't i' ? ' h i J .,7 t, , ,( " [ . -
. ' d'a [s A s"M A' .. -
1" .
-i
. -f , ; J 7'isii , J : M'i ?-
l
? f d . M at:! n k . .
a ERGlii;e 42.8M L.W ? .. ' J u. 5/ A it diId4 l hwia$hl StWWMMHMIMN 58uWniM C. 19 !;@Ilil@@i%d,,%'#4 E. 1PTltip*P"k d'thlA f,g "4* ;"td % V"" r- g -
.- 4 (jgdli j %" M' ;'gji'% 'g,'i'f.1,t 4;1',- D .tw j b) W$i (T4pfW' d ," .'h;. " . , A ?j .
' ~
J
#4
! W W W a " v:t; m n8 4'fy
( 'd ,.M!RM29NN,.i m w a m w .i c t:iit6s w t s c. ;,: g4Y,'$ illH.R'*lWM44 k@ % 440'"/T / "/M de ',L 'd,;u % s .# c n a m?, j i
- b'#
id5@iiANilff ~ M,Mi~,hE.i l"Raikib &w;3 M.64 % & c'b't? ' M B hM h :,T W M W "4 :,i M ,' ',* BBWDE!"'" ARK 'he.h:ieJE "7 %tP'Ja - g
. EMBREEEBRnilisat.Ld.'43"htF' A , . ,
B M E W E BNte m W A G e WA s p: ll l- Ellil59WhM3MiI" 1l T ARi C V a : L-E MedE85tW29aW"a e85Wmow a '
" ' ~ ' -
BWilkd!GibeamitWW' N L&f,mmntn L Egg)gMigig'ig)r W;4if@t.49 ci w o! L E E N G M h 2 M c L * !.iz pi 4 , % ,.a w
- BRGBt!Em s h48.9 %h?M4% n'-
m,u,,r,f m, , &m. -a e=f i,.. ..g m. =, . = . m. .. , M. . . + e. A = . a t59Jund:'8Say.j&g.r : :- : , ' " , - :.
.: , 5. : . . :g . 3. g . _ g . . . .
2 : es . ;m ,3 :. 1 :a _ - P M E M @ @ S M9did i-H:n;1 M M M T W E , 4 *E R 4 E en . ( m. SENt0YAH - tm!T 1 3/4 2 14 en...-e-o.- v+e.,-~-,-,,-. , . ..,..,,,.--.mnen.
*O f' I
?! ;1
( ' POWER DISTRIBUT!oN LIMITS fl i i 3/4.2.5 Se PARAN TER$ . unmNs enamoM Pot oPrRaTroN - 3.2.5 The followiny ONB M1ste,i parameters shall be maintained within 'the . limits shown on Tab'e 3.2-1: ,
- a. Reacter Coolant System T,,,. , ,
C,kgaCbf*b0fontSySM
$PPLICA8ILM: WII Toto l EION b6f6, j EIg: . )
C With s,ny of the above parameters exceeding its limit, restore the parameter to within ittliett within 2 hours er nJuce THERMAL POWER 'to less t.han SE of , , RATE 0 THERH4L POWER within the next 4 hw rs. . t - T C P SURVE!LLANCE RE0UIREMENTS - ' 4.z.s.t b Each of the parameters of Table 3.2-1 shall be verified to be within their ii.its .t i..st per in h. ors. 4.2.;5.2 % Reacir Coo'ont Sptem tohl b rde shoIl be JArmined h be widin h hnet by measurement.#
. least once >er l8 ment)ns. .
s e b e ( SEQUOYAN - UNIT 1 3/4 2 18 .
.G. 46 -
,..m..._.,,._..,....my _ . - , , , _,y-_ ..--,,%.,_w. . . . . __ __
,7
,~
=e
~~
m O- ~ q s 1
- w. VA3tE 3.2 '.
2 - 8 ' N mas.rtats g< . .
~
LIM 1S-E Q 4 Loops In b5 #+ y N Operetten Reacter Coolant Sys, tem T < 583*F
. avg -
i Pressurizer Pressure . ?,2220 psf =* EEndo'r M ot Sy5Icm j 2 378400 3pn; 4 - 3 . ! w. ' A " Limit not appilcable during either a TMUWIM. Pc40t rssy in excess of N it4TED TMDBWIL POWUt l per ninete or a THEIBIM. of servelliance requirement POWUt 4.1.1.3.b. step in excess of 10E RATED THERfML POWER, physics test, or perfetisonce 3 9 - $ e : { 45_ Includes a 3 5 % b measurement uncertoirify
~
f i l . l us . l 4 . I 4 e+ 't
- y . * * \
}
= i f-I
, f e, ,
3/4.2.2. and 3/4.2.3 HE AT FLUX AND nuclear ENTHALPY ! (- HOT CHANMEL VFQ(a) anel FM ., 3/4.2 POWER DISTRIBitTION LIN!TS *
- ' I
- I ,
, BASES <
W specifications of this section provide assurance of fuel integrity ' during events ty: Conditten I (Nemel Operetten) and !! (Incidents of Moderste Frequency) - during fission gas opma(l release, operetten fuel pellet temans in short ters transients, and (b) I te within assumed design criteria.. perature and cladding mechanical properties In addition, limiting the peak linear , power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria lieu of 2200*F is not exceeded. . L I J. ' theseThe definit. ionsare specifications of certain as follows:hot channel and peaking factors an used it ; l FI(2) Heat N1ux Hot Cl.annel Facter,'is deline1 as the maxious local heat flux en the surface of a fuel red at core elevation I divided by the everage fuel red heat flux, allowing for manufacturing tolerances en fuel pellets and rods. I R23 Nuclear Entheipy Rise Hot Channel Factor is defined as tne ratto of the j ogral the average of linear power along the red with the highest integrated power to red power. i' I- : l_ 3/4.2.1 AXIAL FLUX O!FFERENCE fAFD) The limits on AKIAL FLUX DIFFERENCE assure that gthe F (Z) upper bound envelope of 2.237 times the nomslized axial peaking factor is not anceeded ' during either normal operation er in the event of senon redistribution follow-
'ing' power changes. -
provisions for monitoring the AFD en an hutomatic bast's are derived free ' the plant process aceputer through the AFO Monitor Alarm, h computer deter- ! sines the one minute average of each of the OPERABLE excore detector evtputs * '
- and provides an alam sessage immediately if the AFD for at least 2 of 4 er 2 ,
of 3 OPERABLE encore channels are outside the allowed Al-Power operating space and the THERMAL POWER'is greater than 40 percent of RATED THERMAL POWER. "
+ Y 4.2.2 a M
=7 y R M TNA.PilRI5E 44.2.3HDT NEAT CHANN F. FACTORh 0 7 CHANN h FACTOR. RC b LOWRAT .
' \ \ .
The 11mits en heat flux het' channel factor ' and nuclear enthalpy rise hot channel factor ensum that 1). i
/.
1ecal power density and einimus ONBR are not exceeded an of a LOCA the acceptance peak fuel critaria lielt.clad temperature will not exceed the 2200'F ECCS (g ( , SEQUDYAN'- UNIT 1 8 1/4 2-1 December 23, 1982 Amendment No.19 4 8 ___________________._._.______._m--.,...-..,,_.e_. -
- . , ,.,.. c-,,,.,me,. ..m__,.,,,,,,e -.,,.e,.
_ --,,m.-.,-_..--_- _-- ., _ _ . , . , _ ,
t! , ( ,
. c
, POWER DISTR 10UT10N LINITS gg gg ;*
SASES
- T Each of these' asurable but w111 novea11y only be determined periestcally as fed in Specificattens 4.2.2 and 4.2.3. This peristic ~
, a surveillance is sufficient to insure that the limits are maintained provided: .
i a.' Centeel rods in a single group move together with no individual red : insertion differing by more than + 13 steps free the group demand '
,positten.
J
- b. Centre) red in Specifices en 3,1.3.6. e are sequenced with everlapping groups as described :
- c. The control red insertion limits of Specificatter.s 3.1.3.5 and !
f
.. 3.1.3.5 are asintained. ~
- c. The utal power distribution, expressed in terms of AXIM. FLUX 1 DIFFEREhCE, is mafatsined within the limits. 123 ;
Fh will h1M nee wi tts limits med cor.ditio s* 1.hrough
- d. ve are sainta . As reted h Figures 3.2-3 3.2-4, RCS 1 and be traded " against one ther to enau g t the calcu d .
}
4 1 DNbt will be below the ign DNBR value. Tie relaxatie D netfea cf T f Fh as a ' 4 l4WER all neesinthehdialpowera for all i l 7y h pe ssible red ins tion limits. < g l
, When. flow rate a are esasu no addittenal 1ewances are ,
ssary prio comparison, h the lietts Figures 3.2-3 3.2-4. Me rement errors f 3.5 percent RCS totalf1 ate and 4 pers tforFh L have allowed for determination the design value. , L l R lated in Spec, 1 cation 3.2.3 used in Figu 3.2-3, account f F"y,less as ca than equal to 1.4 . This value is value used the vario safety analyse where F ," inf aces'paramete ther than DNR e.g. , peak c1 reture, is the "as measu value all . L Rt . as deft a11sws for the nelusion of a nalty for Red on ONOR Thus, know the *as meas valubs of and RCS flow low for "tr off" in excess R equal to 1. for the purpe of offsetting- Red Sow DNBR nalty. g y,
' December 13, 1982 * '
E SEQUDYMI - UNIT 1 8 3/4 2-2 Amendment No.19 - r b
l INSERT 4 l 1 l l TherelaxationinFfHasafunctionofTHERMALPOWERallows changes in the radial power shape for all permissible rod insertion limits. (gwillbemaintainedwithinitslimitsprovided conditions a thru d above, are maintained. 1 When an Fg measurement is taken, both experimental arror and manufacturing tolerance must be allowud for. The 5% is the appropriate allowance for a fell core map taken with the incere j detecter flux mapping system and 34 is the appropriate allowance for l manufacturing tolerance. i When(H is measured, experimental error must be allowed fer i C and 44_is the appropriate allowance for a full core asp taken with the incore detection systaa. ThespecifiedlimitforPfMalsocontainw [ an 84 allowance for uncertainties which mean that normal operation willresultinFfH $ 1.55/1.08. The 84 allo'mnce is based on the ' following considerations.
- a. abnormal perturbations in the radial power shape, such as fromrodmisalignment,effectFfHmoredirectlythan F.
g
- b. although rod movement has a direct influence upon limitz'g Fg to within its limit, such control is not readily availabletolimitFfH,and
- c. errors in prediction for control power shape detected during startup physics test can be compensated for in FO by restricting axial flux distribution.' This compensation for FfHislessreadilyavailable.
( ~
., . POWER O!$TRIBUT!0N LIMITS *
}!
SAsts penalties app 11ed to y account for Red low (Fi 2-4) as a 4' function e are consistent with described in Mr. John F. 's
, (MRC) letter to . Anaerson (Westinghouse) April 5, 1979 and W 869 Rev.1 (partial red bow ta). .
When measurement is taken, he rimental error and manufa tolerance must be a for. $ percent is the riste allowance for a core map taken with the re detector flux snapping and 3 percent is the riate allowance for menu ritig telerance. ' _ N .h R23 m"
-p .
q Tb het channel fauter Fq (z) is measured periotically and increased by a f\N cycle and hethht cependent pot.or factae', W(t), to pre *ilde assurance that the
,{ iteit on the hot channel factor, Fq (t), is set. W(z) eccounts for the effects of r.ormal operation transients and was determined from expected pqwer contro) '
)
asneuvers over the full range of burnup conditions in the core. The W(t) fvvt'on for norms) oper:rsicn is provided in the Peeking Fsetor Limit Report I per Specification 6.9.1.14.
- 3/4.2.4 00ADRANT POWER T!LT RATIO The quadrant power tilt ratio limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during startup testing and periodically during power operation. The two hour time allowenee for operation with a tilt condition greater ' than 1.02 but less than 1.09 is provided to allow identification ato cor-rection of a dropped or misaligned rod. In the event such action t:as not '
< correct the tilt, the margin for uncertainty on F3 is reinstated by reducing the power by 3 percent free RATED THERMAL POWER Tor each percent of tilt in excess of 1.0.
.}
' December 23, 1982
))
SEQUOYAH - UNIT 1 8 3/4 2-4 Amendment No. 19. e , .--,1 - - - - - - w .- ~ ,,,,,_e_. , ,, -, n, .,n,..s.. ,,--n.
i INSERT 5 l Fuel rod bowing reduces the value of DNB ratio. Margin has been retained between the DNBR value used in the safety analysis (1.38) and the WRB-1 correlation limit (1.17) to completely offset the rod bow penalty. The applicable value of rod bow penalty is referenced in the FSAR. Margin in excess of the rod bow penalty is available for plant design flexibility. l o 1 i
e
. . . t POWER DISTRIBUTION LIMIT $
- I SA$ft l, " ' . ,
( -
)
)
3/4. 2. 5 DNS PARAMETER $
)
i The limits on the DNS related parameters assure that each of the para-
)
meters are maintained within the norsel steagy state envelope of operation . l assumes in the transient and accident analyses. The limits are consistent ' wi % the initial FSAR assumptions and been analytically demonstrated adequate to maintain a minimum DNBR roughout each analyzed transient.
- R23 The 12 hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their i
limits following lead changes and other expected transient operation,
\ greater than or equal to the safety analysis DN8R limit
'k.* i k
. . . 2 L ,
3 December 23, 1982 l [- SE M YAH - UNIT 1 3 3/4 3 5 Amendment Mc. 19 9 ., .
- s. 5 ... m i ., .. .. .
l l .
SUMMARY
OF TECHNICAL SPECIFICATION CHANGES FOR SEQUOYAH UNIT 2 Page Section Description B21 2.1.1 Basis Change W.3 correlation This change reflects B23 to WRB 1 correlation and the DNB correlation B25 added safety analysis used in analyses. B 3/4 2 5 3/4.2.5 Basis DNBR limit. 3/4119 3.1.3.4 - Revised rod drop time 'Ihis change is a result to less than or equal to of changes in the fuel due 2.7 seconds, to the VANTAGE SH fuel design. The effect of this increase on the safety analysis has been considered. 3/42-8 3/4.2.3 FAH and Rod Bow Use of the new rod bow penalty 3/429 Penahy (delete Rod Bow methodology reduces the rod bow 3/4210 Penalty (RBP) as a function penahy. The reduced penalty is 3/4 2-11 . of burnup in FAH equation accounted for in the analysis by 3/4 2 and delete figure 3.2-3). using available DNBR margin. B 3/4 213/4.2.2 and The new methodology is defited C B 3/4 2 2 3/4.2.3 Basis in the references below*. 1 B 3/4 2 4 3/4216 3/4.2.5 DNB parametct This change is a result of 3/4217 the revision of the B 3/4 2 5 3/4.2.5 Basis FAH Tech Specs with respect to flow. Skaritka, J., (Ed.) ' Fuel Rod Bow Evaluation,' WCAP-8691, Revision 1 (Prop), July 1979. Partial Response to Request Number 1 for Additional Information on WCAP-8691, Resision l' letter, E.P. Rahe, Jr. (Westinghouse) to J.R. Miller (NRC), NS EPR 2515, dated October 9,
. 1981; ' Remaining Response to Request Number 1 for Additional Information on WCAP-8691, Revision l' letter, E.P. Rahe, Jr. (Westinghouse) to J.R. Miller (NRC), NS.EPR 2572, dated
- March 16,1982.
Letter C. Berlinger (NRC) to E.P. Rahe, Jr. (Westinghouse), ' Request for Reduction in Fuel Assembly Burnup Limit for Calculation of Maximum Rod Bow Penalty," June 18,1985. l t
L r l
% I
- g. ,'; I 3..
F -. e i: I e I Recommended Modifications to ; 4 the Technical Specifications for -J l Sequoyah Unit 2 1 4 j 1
.J
\
1 t,
% p w: ;
k. O D e = t
~ ,
4 l
'h w
E, k
*9 h
a.
.J e 4 gy- g -
$.1 SMETY LIMTS
{s BASES WRB
- c. w hl scarrth%n u t d e & ord re qW-3e. geeteJapani.e wBl 1 The restrictions of this safety limit prew'ent overtiesting of the fuel at4 k possible cladding perforation which would result in the release of fission ; i
' products to the reacter coolant. Overheating of the fuel cladding is prevented y- !
ty restricting fuel operetf6n to within the nucleate be111pg regime where the ~
. heet transfer esefficient is large and the cladding surface temperature is l slightly above the coolant saturation temperature, S. ;
' y j Operation above the imper boundary of the nucleate boiling regime could vesult in escessive cladding temperatures because of the enset of separture { l from nucleate be111ng (DNB) and the resultant sham reduction in heat transfor coefficient. D e is not a directly measurable parameter during operation and therefore T'tRMA. POWER a tor Coolant Temp re and Pressure have been relates te Jws m raugn t curre " W The DNS It @ hen developed to predict the lux and the lo, cati f or axia Ty unifore and non-unifers heat flux distributions. h local DNS heat flux ratio, DNBR, i defined as the ratto of the heat flux that would cause ON8 at a particular 1-l core location to the local heat flux, is indicative of the margin to DNB.
Mpf N 1he minimum va oftheDN8Rdurinhter.dystateope ion, normal
~
oper aal transients, d anticipated tran ' nts is limited 1.3D. This j, value co nds to a 95 cent probability a 95 percent c(onM4ence level
'that P'AB vi t occur and is osen As an apprort margin to DN r a?1 opersti!p,condi s. N N e
ter Coolanth curves Sysof Figure 7.1-?. eW the loci of points of THERMAL POWER, Reac t assure ar.d avertge temperature for which t.he minimum DNBR is no less than or the averge enthalpy at the vessel exit is equal to R101 the enthalpy of sturated liquid.
.the s,4ty nWo.s M5K WT l These curves are based on an enthalp N of 1.55 and a reference cosine with a peak of 1.5f hot channel factor, Ffor axial po is included for an increase in Fy at reduced power based on the expression: L21 a
Fh=1.55[1+0.3(1-P)] where P is the fraction of RATED THERMAL POWER hse limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control J rod insertion assuming the axial power imbalance is within the limits of the
- f. (delta 1) function of the Overtemperature Delta T trip. When the axial power 8R i
14 balance is not within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits. C . SEQUDYAN - UNIT 2 8 2-1 Amendment No. 21,104 May 5. 1989
-w s- -
v+,g -yg.,yy-m-,c,--w
?
4 C . INSERT 1 The DNB design basis is as follows: there must be at least a 95 percent probability that the miniaua DNBR of the limiting rod during condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 or W-3 correlation in this application). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit (1.17 for the WRB-1 correlation). l' e I l l l l l l
. 1 l
i 4
2.2 LIMITING SAFETY SYSTEM $ETTINGS BASES Manual Reactor Trio ,
. The Manual Reactor Trip is a. redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
-power Ranoe. Neutron Flux The Power Range, Neutron Flux channel high setpoint provides reactor core protection against reactivity axcursions which are too rapid to be protected by temperature and pressure protective circuitry. The low set point provides redundant protection in the power range for a power excursion beginning from low power. The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicate a power level of above ap-coximately 10 percent of RATED THERMAL POWER) and is auto-matica11y reinstated when P-10 becomes inactive (three of the four channels indicate a power level below approximately 9 percent of RATED THERMAL POWER).
power Ranne. Neutenn Flux. High Rates t The Power Reage Positive Rate trip provides protection against rapid flux l increases which are characteristic of rod ejection events from any power level. Specifically, this trip complements the Power Range Neutron Flut High and f.ow t'.'ips to ensure that the critgrjJ are est for rod eiection from partial.pewer. the safety analysis DNBR limit The Power Range Negative Ra p provides protection to ensure that the minimum DNBR is maintained above for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which, when in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control system, could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor for all single or multiple dropped rods. Intermediate and Source Range Nuclear Flux The Intemediate and Source Range, Nuclear Flux trips provide reactor core protection durinn reactor startus. These trips provide redundant protec-tion to the low setpo nt. trip of the bower Range SourceRangeChannelswillinitiateareactortrIpatabout10 Neutron Fluxper counts 4hannels. The second unless manually blocked when P-6 becomes active. The Intermediate l SEQUDYAH - UNIT 2 B 2-3 Revised 08/18/87 l
LIMITING $AFETY $YSTEN SETTINGS ! SA$ES ' Pressurizer Pressure . The Pressuriter High and Low Pre'ssure trips are provided to limit the , pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure , protection, and is therefore set lower v.han the set pressure for these valves (2485psig). The Low Pressure trip provides protection by tripping the reactor 1
<in the event of a loss of reactor coolant pressure. '
Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor
- Coolant System overpressurization by limiting the water level to a volume i sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken for operation of this trip in i the accident analyses; however its functional capability at the specified tripsettingisrequiredbythIsspecificationtoenhancetheoverall l reliability of the Reactor Protection System. i Luss of Flow The Loss cf Flow trips provide core protectiLn to prevent DN8 in the 1
l event of a loss of one or more reactor coolant pumps. ' l Above 11 percent of RATED THERMAL POWEk 3 en automatic reactor trip will ! occur if the flow in any two loops drop below 09% of nominal full loop flow. Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 895 of nominal full loop flow. "
- latter trip will prevent the minimum value of the DN8R from going below T l during normal operational transients and anticipated transients when 3 ll o; are in operation and the Overtemperature delta T trip set point is adjusted to i the value specified for all loops in operation. With the Overtemperature 1 i delta T. trip set point adjusted to the value specified for 3 loop operation, 1
the P-8 trip at 765 RAJT n THERMAL POWER will prevent the minimum value of the-DN8R from going belowIL: hduring nomal operational transients and anticipated transients wita 3 loop 'n operation, the safety analysis DNBR limit i SEQUDYAH - UNIT 2 8 2-5
i
)
' )
REACT!v!TY CONTROL SYSTEMS l R00 DROP TIME
\ 2*7
' . l 1
LIMITING CONDIT!0N FOR OPERATION 3.1.3.4 The individual Yu11 op time from the fully withdrawn position, length (shutdown and control) ri g,ga' shall be less than or equal to seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with: ;
- s. T,,, greater than or equal to 541'F and
- b. All reactor coolant pumps operating.
APPLICABILITY: Modes 1 and 2. i ACTION: a. With the drop time of any full length rod determined to exceed the I above limit, restore the rod drop time to within the above limit j prior to proceeding to MODE 1 or 2. ' b. With the rod drop times within limits but determined with 3 reactor coolant pumps operating, operation may proceed provkied THERMAL i POWER is restricted to less than or equal to 71% of RATED THERMAL ' POWER.
\
SURVElLLANCE REQUIREMENTS 4.1.3,4 the rod drop time of full length rods shall be comonstrated through wasurement prior to reactor criticality: . a. For all rods following each removal of the reactor vessel head, 1
- b. For specifically affected individual rods following any maintenance
- on or modification to the control rod drive system which could affect the drop time of those specific rods, and
- c. At least once per 18 months."
L l t *For cycle 1, this turveillance is to be completed before the next cooldown R20 or by August 5,1983, whichever is earlier.
# Fully withdrawn shall be the condition where shutdown and control banks are at a position within the interval of 1222 and 1231 steps withdrawn, inclusive. I'8 C
SEQUDYAH - UNIT 2 3/4 1-19 Amendment No. 20, 98 March 28, 1989
m . .o. r, ll .. ! 3/4.2.3 MUCLEAR ENTHALPY HOT CHAWEL FACTOR-Fj 4 h (. .. _ rowEn ossmsvrion umis 3.2.3. Fj shall be limked by b 7 _, ' d4.}NB Rh FL h Aft h R 0 N.h! c- . { . LIMIT!ss COMITION 80R OPERATION . .. ){n L 3.3.3 The inatten of icated Rest Coelant Sys (RCS) tota) eu rete and 2 shall 2 intained wi in the regio of allowable h ' I ope tien shown igen 3.2-3 e 4 lose ese tion: .
~ '
\ ,
1 1.49 [1.0 + 0 3 (1.0 - FN g 4,
,. . R. N +- Fu s t.s5 (i.o+o. sci.o-r
_ ti Nur (son . N - Turenn. mwra , RATan TMFllMA: pfhKB 18 . A ~ Fh = Measured lues of Fg obt.ained by ing the able s intere date es to eb n a power d tribution . measured lues of F"g hall be use to calcula Rs e Figure 3. -3. include measurement neertainti
- s of 3. for flew a 41 for i re esasurenegt or Fh a i s ..
x s
- e. R87 (F,u = Led Penalty a furstion f regten avkage -
hornup a shown in igure 3.2'4, where a reg 1 '
. is deft as these semblies w h the same
. N 1eading (reloads) r enrichmeht (first ce .
APPLICA8!LITY: MDDE 1 M: . I h th the tien of RCS tal flow and Rj . evtside regions e1 cceptable ration shown Figure 3. : , Within
- 1. Eithe store the ination of 5 total f rate and Ry , to within above limi er InSerI 2. 2. doce a R and red powst to is the Power than so: e Neutron TED w ux - High tr setpoint less than er ual to 555 RATED THE L .
POWS (thin the t. 4 hours.
*f.
SEP291gs3 - 3/42-8 SEQUDYAH - UNIT 3 Amendment No. 31 i '
; Fi.. . ,:
.. . .._. ~. ._ _ _ _ _ - _ _ _ _ _ _ _
e m ( 1 1
~ '
INSERT 2 ) i With FNB exceeding its limitt
- a. Reduce THERNAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-Eigh Trip Setpoints to 5 55% of RATED THERMAL POWER within the next 4 hours, .
- b. Demonstratethruin-coremappingthat(H is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours, :
and
- c. Identify and correct the cause of the out of limit condition
, prior to increasing THERMAL POWER above the reduced limit required by a. or b. above; subsequent POWER OPERATION may ' ! proceed,providedthat(Hisdemonstratedthroughin-core mapping to be within its limit at a nominal 50% of RATED THERNAL POWER prior to exceeding.this THERNAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours'atter attaining 95% of greater RATED THERMAL POWER.
-~__ _ ___ ____..-- ~...-. . - . . . .-.n....-- - --
(- .; POWER DISTRIBUTION LIMITS .
\(
- Mig: wtinued) ,
b.- WitA1 4 hours e nitially bet outside above limit verify 1 through ere flux ing and tal flow te esaparise that . the ina .n .f R,, .nd Res to fi ut. motor.d ; thin the abo limits, e reduce.TNE POWER to s than SE o ; RA THEML within next 2 hours. Identify nd correct cause of out of-Its conditten for to increas THERMAL above the N duced,THE POWER lia - stred by A 011 items a. and/or b' ove; subsequ t POWER OPE 10N any p eed provided t the e ination of Rg , Rgand
. indica RCS total low rate are emonstrate 'through inco flux
.mSPJi 3 mapping a RCs total low rate coup ison, to within the re ion
, of acceptabi operatl ion l on Figure .2-3 prior o exceeding
, lowing THE POWER lov s: ,
6 4 1.
, inal SOE o RATED'THE L ?OWER.
t 2. A
+ (nal 755 of D THERMAL 'R. and
- 3. Withih hours of at ning great than or equa to 95% of RATED THE L POWER. .'
SURV LLANCE . REQU! EMENTS .
. N i
. . s .
s -
\ \ \. ,
N .2.3.1 7 revisions e Specifica\t n 4.0.4 are t applicab .
- 4. .~ 2 The instion of icated total flow ra and R g , ,shall be date ned to be in the reg n'ef accep le operatio f Figure
-3:
y a. 1 ior to ope tion above ing, and of RATED RMAL POWER ter each f 1 .
- b. At le t once per 1 Effective 11 Power s.
a SE @ M ~ UNIT 2 3/4 2-9 ' 4 e e
4 e
- i r
w n ScrE .3 (. l 50RVE!LLANCE REQUIREMDffs 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. l 4.2.3.1 F"gshall be determined to be within its limit by using the novable incere detectors to obtain a power distribution sep:
- a. prior to operation above 755 of IIATE THDDUd. POWUt after each fuel '
, -leading,and *
- b. At least once per 31 Effective Full Power D es.
- c. The esasured Fh shall be increased by 4 for measurement uncertainty.
l 9 9 9 9 P L
. , _ . _ , - + - - - . . _ - , , .,.. _ . _ - .- , . , , . . , - ,.-..I- - - - - , -. ._ -_ _
r i i
. 1 POWER DISTt!8UTION LINITS SURVE!LLANCE REQUIREMENT $ (Cont {nued)
/ 1
\
2.3.3. The indicat 5 total flow a shall be v fiedtobhe hin the l- n of acceptable oper the ao recently obtaines va n of Figure . -3 at least e 1 per 12 he when ; tion 4.2. . are assumed to ext s of Ryand 2 obtained per ifica-3.4 The RC tal flow rate indica CALI T!ON at less a shall be a ted to a L ce, per 18 months,
- 4.2.3.5 The $ total' flow s, shall be deteraf by measuremen t least once 9 r 18 S
I G e l 6 4 4 C . r
. g e 4
9 4 4 9
,t 9
4 O 9 0 4 9 9 g 4 5 S 4 0 4 i II N M
- W II I
)[(g.10 Q
4 6 P e
, 8 9 9 9
f ((-
\ '
W g,
- :=a=ct;;a eO _
M E.
. a $
t=2= .* 5
= y a + t' s -
E c : _, igs
. = - -
..; _ p b
.. g; a:; 3
' __ a
~~ N -g
< = , .e '
==
mm
==
wmmW= ~
-..!*a*E'in-4 ge
- l e
89 _=, .-. . g
=
a . __ _ar. . - _ _ _. = 1,52 g3e {t
._ -===
-4.<
i y ---. c; m se
- p -
U*
- 2. 5 8 ' -
w= __ i 5 ,.,E . C( g .<x - er- -
" .t -
8 __ _ i p. _ e g 2a =. _.-
- c g
~
- ,_ _ 55 l, - -
._.g l
; B -?
* = Z ._.
A - m=
- gi E
\ .go d u -4 a. . = - .i, n w 1 *; > --- yss:= _ ', - n= g I;;p =.- sim E _ jags =sa_ a_ a Em 1
.EE edendesiigi t._ 1 WT=__=E -nE m a i
Ag a w 4 8 D
. 'R -
1 l
\
l l puo,ou aAvumou wAoA sou l l 1 i
'SEP 2 91983 SEQUDYAH - UNU ? 3/4 2*11 1 Amendment No. 21 l C l l
l
. l
. . .. j
....) .,,
l 1 c .
, .. =- .. .=
_=_....~... p il [r :' ' l .5~ .::
. ~t k ( 4 I " :~ l . .. : "._i l
.a. .,-y :=e F > 7 ,
r+1,
; - . z, - -. . . .
A b .g g c w ;- V-1: - r' _1 . ..
~
-s ,
z ---=u- ,e = - x -
, . . k - -- : ,i , .u 1 1 : .
.. :.1=
1
..-u_
e,;tD.-- 64.W E EJ'! ? d,, 's 1 : :a w 1 . _. - .: . l i
' -' i R ~ :-' -
;? s'T -
)
c ' i 2: :=>g . efj % g=i7 h:j:,+ g ,h E .
- t i-- - - ;
G= ~.. M u M M Ev % . , _ ~ ..igi .. --: x . , =_ L C # =ii
^ } 1 } g ., "'. ^' f. 2 . ' ' ' '.
7 . ~ . .. ' ' _=_ . . - . '_.?.
#35,d] 4: .~il 4-1.: ; r,, : ; I z - . - . ...: .
-H l ; .
I b Q t n w =? g 1._'==E =i ,; E; = 1; - .. ;.?: =y : Ah ! g StyE:L- $ E2iE. r,: r .E 5g5-?;p 7:@= 2. 5 = 'i::s'- i :i= = '3 i=- o 2
= 2 ----
. :-- +:5 5 i- i 7 . :
a t r n- - -- - r- : YWN&hkW h lh?i?5$555 i h , BEttilEBH5GlEM515;iRL%E452MMkiW mEmemamammewereeswa.w [. ) 1 MBEEMMGEMiliHWJits@&w2C;iSe , 5 u R E ESEEiE! BERM!tth @ hiialitTA22 su W E W !!iB81dMR NW5HiiEitAMc! @-W;EVs E* sf '
.Y # mig E202EIW2Hi!"P;N!MPMMi$s=h 5
E J EgpJ:ineg=gsge;Apig== 'wp@tf: 4 8 ( o/ RtgiEJ5?=4ami=#=eEsMi!E=*mnH - D I. Q B M Bi M E:k!!M M M #J 32i,t @ @ 4 esW y
%,l 525ilMB224222iisaFEinE!@es9tMy < ,
L
'!Eigl@!Hinsidii"R322W= ~9:iiWisis-itz g ElHEliWhi5e BB 1hi @"r=s;RCleig E l !
L. 15iBgaggg __ $ .
~ = E _a s qms lg f .
EE = EG5MG55 f dWW35W , 2
@Etl45iidt:22issiWWC$i=@ithe!
gg:_ . nw:_E =_ g==- - aac === =E=
--5 am ? '
miB M G . Mhig4f i!jphiM M ($!* > M . M W G t p s y g n W i4 MSt&AJA WEni EiM56 ' L t N h d5 E 5 b M, 4 s i M EF 55 asi E E: =. ENNENbMNINEW
~ = . _ = _ = = = = = = = . = = . _ = = = = =+.=
l- ' E.Ex sEiEEE EE 5 E EEiZZ EFE E E E 52 E =~ E- E E EE. r}E-. * ~-~~ EE EE Z Z: W g 5.-g.M.r_T. e 4 4 4 4 E
- o NW I: .
1 SEQUDYAH UNIT 2 3/,4 2-12 L . 1 l
- i. .
~
POWER DISTRIBUTION LIMITS
" / '
3/4.2.5 De PARAPETDts .
, - 1
.' LiWiiiNR COMfTION FOR OPERATION * '
3.2.8 The fo11swigi 8 5 related pareerters shall be maintained within the Iteits shown en Tab'e 3.2-1: . a. teactor Coolant System T ,,. - T""' *" '"* *"" c, Reactor Coolont Sysh APPLICABILITY: ICDE 1. Tota plow Royg M: . . trith any of the above parameters exceeding its limit, restore.the parameter to . within its limit within 2 hours or redu.e THERMAL POWER to less than SE of , SATED THERMAL POWER within the next 4 hours. .
$URVEILLANCE REOUIREMDf75 -
4.2.6.I l i M Each of the parameter; of Table 3.2-1 thall ha verified to be within n twr usit. at ie.st e e ,vr u hoves. >
'4 2p z& Reoar Coo'ont spfem fot I Clow rate shoIl be L determmd h be winM in lim;f by meosurment.#
.. leost once nr 18 rnends. .
e 6 l Eb g I
,-,e .---, ,, .~,....--- . , . ..
. u:. = -:
~ -
p3 . f ,
~
s .
.t .
se .- . g, a TABLE 3.2-1-- .
;j- .
DNS *ARWETERS *'
~
^
_7
-e LIMITS. '
ee 4 Loops In ~ ! PARAMETER ~ l . Operat1on W.. 'i
./ J'~,
Reacter Coolant System T < 583*F avs - . Presserfrer Pressur# > 2220 psfa* t
.g Reador Co. lor.t System a 3784oo3pm_a e .
>* t
[ i 4
)
t l i 4
~
i .. . . p - i l eM .
.g
}".
- ay
W* i , oG
- Limit not app 1fcable during either a Ti4EB94L P0ldER reap in excess of SE of RATEC THEIWmL i
w3 POWER per minute or a Tf8ENMAL POWEst stee in excess of 10E of RATED TIElgmL POWER, physics test, or perfenmence of, survefilance reepwireskat 4.1.1.3.b. i l # Includes a 3 5 % be measummc4 unartoink.. _l. _ _ . . . . . . . _ . . . _ . . . - . - . ~ . -' _-_ l=
t
.. 3/4.2.2. and 1/4.L,5 flEhT FLUX hMD MUCLEhR ENTHALPY
< , NOT CHAN AlEL -FeCE) and.FA
.((
3/4.2 POWER D!sTRIBU110N LIMITS
.AsEs .
The specifications of this section provide assurance of fuel integrity
$. during events by:
Condition I (Normal Operation) and II (Incidents of Moderate Freque ,3 during norma (l operation and in short ters transients, and (b) limJ - fission gas release, fuel pellet temperature and cladding mechanical properties - to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial 1 1 conditions assnmed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded. 1 1 ' The definitions of certain hot channel and peaking factors as used in these specifications are as follows: g gi FI (Z) Heat Flux Hot channel Factor, is defined as the maximum local l heat flux on the surface of a fuel rod at core elevation Z divided
'by the average fuel rod heat flux, allowing for annufacturing
! tolerances on fuel pellets and rods. .
- F" Nuclear Enthalpy Rise Hot Channel Factor is defined as the ratto of the (t -
1Nagral of linear power along the rod with the highest integrated power to the average red power. - g4.2.1 AXIALFLUXDJFFERENCE(AFD) ' The 11 sits sn AXIAL FLUX DIFFERENCE assurt thkt the F (Z) upper bound g l envelop 6Lef 2.237 times the normalized axial peaking far. tor is not exceeded [ during 11ther normal operation or in the ev'nt e of manon redistribution follow-ing power changes. - Provisions for monitoring the AFD on an automatic buts are derived from ' tha plant process computer theru)4 the AFD Noritor Alats. The soaputer deter-eines the one utn;te average of each of the OPERA 3LE axtore detector outputs and provides an alare message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are eutside the allowed AI-Power operating. space
' and the THERMAL POWER is gre6ter than 50 percent of RATED THERMAL POWER.
AM.2.2andh4.2.3 HEAbUX H0T C L FACT 0 5 FLOWR AND NUCCEAR ENTHAL.P KRI5E NOT C" G EL FACTOR \ \ A.
- The limits on heat flux hot ' channel factor, ENL wd and nuclear enthalpy rise hot channel factor ensure that 1) the design faits on peak
local power density and etnimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS
. acceptance criteria limit.
SEQUDYAH = UNIT 2 S 3/4 2-1 Amendment.uo. 21
- SEP 2 0 383
[,
, ]
POWER DISTRIBUTION LIMITS . Of C % h0C. hoCIorS drd. - C , SASES
)
I
, , e ,
\
Each of these measurable but will nomally only be detemined periodically as fled in specifications 4.2.2 and 4.2.3. This periodic . surveillance is sufficient to insure that the limits are maintained provided:
- a. Centrol r6ds in a single group move together with no individual rod insertion differing by more than + 13 steps from the group demand position. ,
1
- b. Cont ml red groups are sequenced with everlapping groups as described
.in specification 3.1.3.6. - '),
o
- c. The control rod insertion limits of Specifications 3.1.3.5 and '
l 3.1.3.6. are maintained. I
- d. The axial power distribution, expressed in tems of AXIAL FLUX DIFFERENCE, is maintained within the limits.
Fhkillbe ' ntained with its limits pro ded condition . through
- d. ve are maintai As noted o Figures 3.2-3 a 3.2-4', RC$ '
^ be " traded of against one ther to ensure t the calcu ted andFh g,_
R will no below the de DNBR val The relaxation fFhasa ); fu ton of THE L POWER allows nges in the dial power shap for all pemis le rod inse f ora limits. '
'h When RCS teto and are measu ep ddditionalylowances tre kssary prior to arisenwl% the limits of urds 3.2=3 - 3.1-4.
, Meas) nt errors of percent fh RCS total flhw r e and 4 pere tforFh ;
have been 110wed for in Prination t% design DN8 al ue. L Rj , as cale tad in Spacift ton 3.2,3 a used in Figure .2-3, account for y ess than or ual to 1.49. s value is e value used in N various saf y analysts mF g influe parameters ther than DNBR, g.- l - peak clad ture, and is the maxi "as measure value allowed, as defined, al for the i usion of a pen y for Rod on DN8R only. Thus, knowing "as measu RCS flow ow for valuesofFh
" trade o " in excess of R ual to 1.0 e the purpose o offsetting Rod Sow DNBR pe . -
{ SEQUDYAH - UNIT 2 . e 3/4 2-2 SEP 2 91983 Assadment No. 21
~.. . - _ - _ - - . - - - - - - - - - . - . . - - _ _ _ _ _ - - _ - _ _-. I l ( .
^
INSERT 4 TherelaxationinP[gasafunctionofTHERNALPOWERallows changes in the radial power shape for all permissible rod insertion limits. (gwillbemaintainedwithinits,limitsprovided conditions a thru 4 above, are maintained.
]
When an Fg measurement is taken, both experimental error and j annufacturing tolerance must be allowed for. The 5% is the l appropriate allowance for a full core map taken with the incore j detector flux mapping system and 3% is the appropriate allowance for J manufacturing tolerance. When(Hismeasured,experimentalerrormustbeallowedfor ( and 44 is the appropriate allowance for a full core map taken with the incore detection system. Thespecifiedlimitfor(Halsocontains an St allowance for uncertainties which mean that normal operation N will result in F g 5 1.55/1.08. The 8% allowance is based on the following considerations. - a .. abnormal perturbations in the radini power shape, such as -
.from rod stualignaant, effec-t PfH more directly than '
I. g .
- b. although rod movement has a direct influence upon limiting l Fgto within its limit, such control is not readily
, availabletolimitF[H,and l
- c. errors in prediction for control power shape detected during l
startup physics test can be compensated for in Fg by restricting axial flux distribution. This compensation for FfHislessreadilyavailable. L l
3 g.a ' - ' Jh[ ;.; '
- ,y .'
- y. -
QVP . POWER O!$TRIBUTION LIMITS . h eg BASES
. /j I
Sfi? , , .[ nalties applied to g ounEforRodRow(Figure , as s'
~ function of are consistent with those ribed in Mr. John F. Sto :
(NRC) letter to T. . rson (Westinghouse) date il 5, 1979 and W 8691 - ?
, .-lipartialrod, bow'tes . .
When an Fg roment is.taken, both exp ntal error and manufactur n tolerance must be all 5 percent is the approp allowance for a ful map taken with the inco ctor flux mapping system 3 percent is the approp al'1owance for manufactur lerance. . w [ M 2nScr The het channel-factor Fq (z) is measured periodically and increased by a i 5 *rc' ' f ad ** ' '"' ""a*"* "* r '**** r "( * ) ** 'r' ** ** *"r'ac' **** 'h' limit on the hot channel factor, F (z), is met. W(z) accounts for the effects- , q , of normal operation transients and was determined from expected power control a maneuvers over the full range of burnup conditions in the core. The W(z) ;
- function for normal operation da provided in the Peaking Factor Limit Report )(
L per specification 6.9.1.14, 3/4.2.4 00ADRANT POWER TI. IO The quadrant powa ratio limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation. The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and cor-rection of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on Fn is reinstated by reducing the power by 3 percent from RATED THERMAL POWER fur each percent of tilt in
- l. axcess of 1.0.
L s . L( A. SEQUDYAH - UNIT 2 8 3/4 2-4 SEP 2 91983 Amendment No. 21 y
a; > S INSERT 5 l Fuel rod bowing reduces the value of DNB ratio. Margin has been j retained-between the DNBR value used in the safety analysis'(1.38) and ! the WRB-1 correlation limit (1.17) to oceipletely offset the rod how l penalty. The applicable value of rod bow penalty is referenced in the FSAR. Margin in excess of the rod bow penalty is available for plant design
. flexibility.
+
9 l 9 4 e 4
~
l C
I i i y _,.
. l
. POWER DISTRIBUTION LIMITS
[ CW BASES 1 ( f. i ' l p, . 3/4.2.5 DNS PARAMETERS - L l The limits on the DNB related parameters assure that each of the para- J
- esters are amintained within the normal steady state envelope of operation assumed-in with the initialthe transient FSAR assumptions and-accident and analyser. -The limits are consistent
~ adequate to maintain a minimus DNBR been analytically demonstrated roughout each analyzed transient. '
1 g. The 12 hour periodic surveillance of these parameters through instrument K '
. readout is eufficient to ensure that the parameters are restored within their. '
limits following load changes and other expected transient operation. l
' greater than or equal to the b safety analysis DNBR limit ,
L , 4 i I
?
e C d . p 5 {' SEP 2 91983 . SEQUDYAH - UNIT 2 8 3/4 2-5 Amandaant No.11
- ~ ..._... ._.. -
. . Ia
-( ' T
' i t
- t 1
T 4
, APPENDIX B Significant Hazards Evaluation ,
,,~
' 1 i i 7
k i ( . c 4 1 i .-\-
.^i-
- - - . . . . . . ~ . _ . . . .
V SIGNIFICANT HAZARDS EVALUATION
~
he Tennessee Valley Authority (TVA) plans to refuel and operate the Sequoyah Nuclear Plants
- with . Westinghouse VANTAGE 5 Hybrid (V5H) advanced fuel product features that incorporate i low pressure drop zircaloy grids and Removable Top Nozzles, Integral Fuel Burnable Absorbers ,
' and extended burnup capability, his upgraded fuel will also contain Debris Filter Bottom Nozzles, ,
snag resistant grids, and standardized pellets. Dese features have been implernented in other ; Westinghouse reload cores. ' In addition, the Plant Safety Evaluation for the Sequoyah Fuel Upgrade implements the current rod bow: methodology to reduce the rod bow penalty described in the Sequoyah Technical Specifications. .+ _j Finally, the evaluations performed for this fuel upgrade also accommodate effects from the following programs:
- 1. RTD Bypass Elimination
- 2. Eagle 21 Digital Protection System
- 3. Upper Head Injection (UHI) Removal '
4; Boron Injection Tan (BIT) Removal ,
- 5. New Steamline Break Protection ,
6; Low Feedwater Flow Reactor Trip Elimination
- B. Pronosed Channes
'As a result of the fuel upgrade, the following changes are proposed for the Sequoyah Unit 1 and Unit 2 Technical Specifications:
- 1. Modify the Bases for Safety Limits to change the W 3 correlation to the WRB 1 correlation and to revise the associated design Departure from Nucleate Boiling Ratio .
(DNBR) limits.
- 2. Modify Specification 3.1.3.4 to incorporate a new rod drop time of less than or equal
- to 2.7 seconds;
- 3. Modify Specification 3.2.3 to delete the Rod Bow Penalty as a function of burnup in the FAH equation and delete Figure 3.2 3.
- 4. Modify Table 3.2-1 of Specification 3.2.5 to define the DNB-related Reactor Coolant System (RCS) Total Flow Rate limit, including uncertainties, to be 378,400 gpm.
~ . . m s ~. - -- = i_... s..& ..w._
f l' o 3 ' i 9 ; C. Reasons for M.r- =a rhnnes C R Changes 1,and 2 are required to allow implementation of the improved fuel design for
. Westinghouse V5H fuel.
~ Change 3 is required to incorporate new evaluation methodologies for the effects of fuel rod bow on DNB. 'Ihe new methodologies provide a basis to eliminate unnecessary power distribution penalties and to simplify the specification.
.7
~
Change 4 is required to relocate the RCS Total Mow Rate requirement from Specification 3.2.3 to 3.2.5, as a result of Change 3, and to clearly" define the DNB Gow parameter limit. This limit includes Dow measurement uncertainties. Overall, the proposed changes for this License Amendment Request (LAR) are the result of three -- primary differences: b . 1.
-Increased rod drop time due to the reduced guide tube diameter for the V5H zircaloy L . grids -
- 2. The use of a new DNB correlation
- 3. . Incorporation of the current methodology to assess the rod bow penalty D. Justification for Pronmed Channes As discussed in the safety evaluation for the fuel upgrade, the previously reviewed and licensed ;
L ' Safety Limits for Sequoyah are met with the upgraded fuel. Tne new fuel design has provided satisfactory operational performance in fuel assembly demonstration programs since the early 1980's. The V5H fuel is both mechanically and hydraulically compatible with the current Sequoyah fuel L assemblies, control rods, and reactor internals interfaces. [ The V5H fuel satisfies the current design bases for Sequoyah and it meets design requirements for hydraulic stability and structural integrity under seismic /LOCA loads, with margins comparable to l 17x17 STD fuel assemblics. Nuclear characteristics are comparable within the range normally seen from cycle to cycle due to fuel management effects. No change in fuel rod design criteria, methods, or model are necessary with transition to V5H, with
. the exception of a new DNB correlation. Based upon the information provided in the evaluation, the Sequoyah plant operational limits will be satisfied with the proposed changes.
-e
r . J ,Y -{: . he evaluation' considered the effects of the proposed Technical Specification changes on the C ' following areas:
- a. ' Mechanical, Nuclear, and Thermal-Hydraulic Fuel Assembly Design .
L b. Non LOCA Accidents , I
- c. LOCA Accidents
- d. Environmental Consequences of Accidents -I There ar:as have been evaluated for the impact of all proposed changes in this LAR, including the transition core effects (with a mixed core fuel loading with both V5H and 17x17 STD fuel). The required analyses as described in the fuel upgrade evaluation were performed by Westinghouse 4 using methods and procedures previously approved by the NRC.
- 1. DNB Correlation Channe (Channe 1)
The calculational methods currently used for 17x17 STD fuel assemblies and described in the Sequoyab PSAR are applicable to V5H fuel assemblies, except for the DNB correlation. The new correlation basis for DNB performance is the WRB 1 correlation. De WRB 1: correlation establishes a DNB limit which provides for the margin of safety i
! required by the current FSAR (i.e., DNB will not occur on at least 95 percent of the limiting
( fuel rods during normal and operational transients and any transient condition arising from faults of moderate frequency at a 95 percent confidence level). The WRB 1 correlation takes credit for the significant improvement in the accuracy of the critical heat flux predictions over t previous DNB correlations. 2.1 Increased Rod Dron Time (Change 2) The V5H fuel design incorporates a snag resistant, low pressure drop, zircaloy grid, ne Zircaloy grid will provide for an enhanced performance relative to the current Westinghouse 17x17 STD fuel product. < Utilization of zircaloy as a grid material instead of inconel reduces the source of cobalt in the core. Consequently, radiation fields due to the transport of activated cobalt should be lower. He snag resistant feature results from outer grid straps which are modified to reduce the potential for grid damage and assembly hang-up from assembly interactions during fuel assembly removal. De zircaloy grid also contains features that minimize hydraulic resistance. 9 _ ___m-__---.__--___m___ _ - _ _ _ - - . - _ _-_ - y
- . .~ ._ -
c, P' g , 4 4 pp,' *
- In order to maintain' mechanical compatibility between the V5H grid and guide tube, a f
( ,( reduction in the V5H guide tube diameter was required. De allowable rod drop time of Specification 3.13.4 must be increased due to the increa=ed dashpot effect resulting from the. ' e guide tube diameter reduction. [ l 3. Elimination of Rod Bow Penalty (Change 3) Fuel rod bow has been observed in Westinghouse cores. De phenomena of fuel rod bowing 3 must be accounted for in the DNB safety analyses of, Condition I and II events. The current ; licensing basis offsets the DNB effects of rod how by partially accommodating it with margin in the W.3 correlation.- The remainder of the rod bow penalty is applied as a penalty on the-FAH Technical Specification. New statistical methods have been developed by Westinghouse which verify that the past treatment of rod bow penalty provided an overestimation of the effects on DNB. Application
^'of the new methods to Sequoyah for the Standard and the V5H fuel products has verified the reduction in rod bow penalty. The reduction allows for accommodation of the entire penalty in the establishment of the Safety Limit DNBR.
l; The requested change to Specification 3.23 will remove an unnecessary power peaking penalty . l and simplify the format of the Specification. The current specification format includes reactor L coolant system flow not only to maintain the minimum required RCS flow but to provide for L an additional offset .against rod bow penalty. The use of reactor coolant system flow to compensate for rod bow penalty will no longer be required.
. The. flow limit (and its associated uncertainty factor) as it relates to DNB will be moved to Specification 3.2.5, DNB parameters (see the discussion for Change 4). Similar Technical Specifications changes related to rod bow penalty have been performed for Farley 1/2, North Anna 1/2, Beaver Valley 1, and Salem 2.
i' ! 4. Definition of DNB Parameter RCS Flow Limit (Change 4) [ ' ne RCS flow limit and its associated uncertainty factor have been moved to Specification 3.2.5, DNB Parameters, which now established a minimum allowable r'enctor coolant system l (RCS)~ flow to prevent violation of the Safety Limit DNB during normal operation and
. accident conditions.
The minimum flow rate is based on a thermal design. flow rate of 365,600 gpm plus the application of a correction for measurement uncertainties (minimum flow rate = 365,600 gpm x uncertainty factor). The uncertainty factor of 3.5% is based on flow measurement uncertainties and feedwater venturi fouling. Herefore, the RCS flow limit for the Sequoyah units is 365,600 gpm x 1.035, or 378,400 gpm. d . _ _ _ . _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ . _ _ w
2
)
y p
<l E. Determination of Sienificant Hazards Pursuant'to 10CFR50.91. TVA has determined that operation of the facility in accordance with the
. proposed LAR does not involve any significant hazards considerations as defined by NRC ^ ' ' -
regulations in 10CFR50.92. - De following discussion describes how the proposed amendment satisfies each of the three standards of 10CFR50.92(c). Operation of the Sequoyah Units in accordance with the proposed Technical Specification changes:
- a. Does not involve a significant increase in the probability or consequences of an accident previously evaluated.
1 The_ evaluations of the mechanical, nuclear, and thermal-hydraulic design effects support the conclusion that the requested changes are within the current design criteria established in the FSAR. Consequently, no new mechanisms have been introduced to increase .the probability of a previously analyzed accident occurring. The accident evaluations (both LOCA and non LOCA) exhibit results which maintain the confidence icvel in the physical integrity of the fission product boundaries as defined in the FSAR.
%erefore, the consequences of the accidents do not increase,
- b. Does not create the possibility of a new or different kind of accident from any accident
- previously evaluated.
( The evaluations' performed establish that the FSAR design criteria and system responses during normal and accident conditions are bounding with respect to the proposed
' changes. The changes will not affect the function of any protection system and they will not introduce hardware which is different in design criteria requirements. Derefore, no new mechanisms have been introduced that would create the possibility of a new or different kind of accident from those previously analyzed.
- c. Does not involve a significant reduction in the margin of safety.
The evaluations performed by Westinghouse addressed all design criteria and accident analyses. In performing the evaluations, the Safety Limits established by the FSAR and Technical Specifications were not modified such as to reduce the difference between the Safety Limit and the limit defined as the failure point of a fission product boundary. Therefore, the margins which were assumed in the accident analyses remair bounding for the proposed changes. F. Conclusion The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists, his guidance (51 FR 7750) includes examples of the types of amendments that are considered not likely to involve significant hazards considerations. -
m 4
.o .I j
i- , .m p i' ui .
+ t Based on' the ' evaluation summarized above, TVA has corsluded that- the proposed Technical k
C - Specification changes correspond to the examples in 51 FR 7750 for Amendments Considered Not - Likely to involve Significant . Hazards Consideration. - Additionally, the proposed changes are
', consistent _with the requirements of 10CFR50.36 and 10CFR50.59.
nj., , s _ks' f s 6 4 4 i L s eg,
= .
t* g O 6 .
, v w ,-. - - -- a m e ,m.-c
D, ,;.
~, . .i ; ,
9 i'.;7 Y 4
,r
)*
- h. .
t ' 1 i
. . ..m
'_h',
- a-o .%
3; . APPENDIX C.
- Nuclear Safety Evaluation Checklist-
'I
.v s
1
)
i 1 t
'?
b l 1 l: (. 9 l l l 1 e i l l l a. i
- i l
l l. i ll l i
y . p
.SECL 89-1049'
- 4t X , Page 1 of 2 Westinghouse Reference No(s).
WESTINGHOUSE NUCLEAR SAFETY l SAFETY EVALUATION CHECK LIST
- 1) NUCLEAR PLANT (S) SE000VAH UNITS 1 & 2 _
- 2) CHECK LIST APPLICABLE TO: V5H FUEL UPGRADE
+,
F l
- 3) The written safety evaluation of the revised procedure, design change '
p , or modification required by 10CFR50.59 (b) has been prepared to the-extent required and is attached. If a safety evalut. tion is not required or is incomplete for any reason, explain on Page 2. ., L i L 1 Parts A and B of this Safety Evaluation Check List are to be completed i only on the basis of the safety evaluation performed. CHECK LIST - PART A - 10CFR50.59 (a) (1) i L . Yes )L No A change to the plant as described in the FSAR? l l . Yes No X A change to procedures as described in the FSAR? ! b . Yes_ _ No X A test.or experiment not described in the FSAR? ( . -Yes_1;No . A change to the plant technical specifications? (See Note on Page 2)
- 4) CHECK LIST -- PART B - 10CFR50.59 (a) (2) (Justification for Part B 4
, answers must be included on page 2.) ;
(4.1) Yes _ No jk Will the probability of an accident previously
. evaluated in the FSAR be increased? .
p (4.2) Yes No ]L Will the consequences of an accident previously j evaluated in the FSAR be increased? (4.3) Yes No jL May the possibility of an accident which is different than any already evaluated in the FSAR L be created? L (4.4) Yes No.jL Will the probability of a malfunction of equipment L important to safety previously evaluated in the 1 FSAR be increased? (4.5) Yes No 1 Will the consequences of a malfunction of equip-ment important to safety previously evaluated in the FSAR be increased?
- l. (4.6) Yes No ]L May the possibility of a malfunction of equipment
- important to safety different than any already evaluated in the FSAR be created?
(4.7) Yes No_]L Will the margin of safety as defined in the bases to any ' technical specification be reduced?
9
+
,%f C :: / r*+A' f
e~A
.s J f 9g P.,
4,# e vI d*"s
/~
e/sp*,
$e t hA:f*k'A/
U e ,$p re ** /e/yg if k' -
$ s /
/
f4s /+ # p o [/ 0 g,s e / /
# e./ #:e s
,,s/ll+&ss
- t p $*/s.
4 .- a. f / c# $ .
/-
+ sl # +,//o/a/ #a~
+
** c 4/.
a
. s* :<+ /
/
/. '
f.f.l 'f
+
c.
/O e \\ \Vs-J
SECL 89-1049 Page 2 of 2 If the-answers to any of the above questions are unknown, indicate.
~
under 5) REMARKS and explain below. ' l If the answer to any of the above questions in Part A (3.4) or Part B cannot be answered in the negative, based on' written safety , evaluation,.the change review would require an application for license !
- amendment as required by 10CFR50.59 (c) and submitted to the NRC pursant to 10CFR50.90. ;
~5) REMARKS:
The following summarizes the justification based upon the written safety ! evaluation (1) for answers given in Part A (3.4) and Part 8 of this Safety Evaluation Check List: SEE OTHER SECTIONS OF THIS PLANT SAFETY EVALUATION REPORT
; .(1)-Reference to document (s) containing written-safety evaluation:
FOR FSAR UPDATE Section: Pages: Tables: Figures: FSAR MARK-UPS: CHAPERS 4 & 15 (To BE FORWARDED UNDER SEPARATE COVER) TECHNICAL SPECIFICATION MARK-UPS: SEE APPENDIX A 0F THIS REPORT Reason for / Description of Change: ( SEE OTHER SECTIONS OF THIS PLANT SAFETY EVALUATION REPORT M D E: h i t . z o . 73 Prepared by (Nuclear Safety): L.V.Tomasic Date: Coordinated With Engineer (s): Sionatures_On File Date: { Coordinating Group Manager (s): Si on e Date: Nuclear Safety Group Manager. . re h Date: //! Meg
~ ~ ~ ~ ~ ^ ~ ~ ~}}