ML19327B405
| ML19327B405 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 10/19/1989 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML19327B398 | List: |
| References | |
| NUDOCS 8910310109 | |
| Download: ML19327B405 (200) | |
Text
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has iW . i> 8*E TABLE 2.2-1 (Continued)
" x:
l 89 *r TABLE HDTATieII5 (Continued) , j 98 8 t ' om IIOTE 1: (Continued) um i g5 3
- t. = Tlas constant utfifred in the sensured T 1eg compenseter, to = 9 s, h53 T' lcg ] < 508.4*F (Nemine1lT et IWITED THEIWWIL POWER),
L
** " 8 85134 Rs
! 'O P = Presserfrer pressure, psig, 1' P' = 2235 pstg (Nomine1 HC5 sperating pressure).- 5 = Laplace transform operator, s 8 ! and f (&I) is a functlen of the indfcoted difference between top and bottom detectors of the ! ! '? power-range neutvwn den cheebers; with gefes to be selected based on measured fastrument i
~
~~$ KifyE'O f M E N ),*."I'sn's N.N(un.II Cackse.Je4; o.M 1 c de o 2 avJ an':rh (1) for g g between -.s and +1eV f,(as) = o, where g and g are percent '
l nnTEn inElegit FouEn fa the to, one hette. helen of the c.re pecstwir, and g + g fe j total THEIWWIL POWER fa ercent of RATES POWER- - _(UvM I CagdeIa.4t)h3 2C@p),am.) e s3%( UentI C3 3 m w4 M , O n d 2. M e 2 j a Q N , c -
!!she 91 A by .
h'b sii j ! NOTE 2: The channel's menteum Trip 5etpelat shell not onceed Its computed Trfp 5etpefst by more then l ( 3.ss .f af s,en. . (nQ f each perced Li ne tw.gn.14 ef g4.-16 emed 3 -32.%, tie. 4T % t. . selpoM shall be c.biadi6dh reducd Lg1.o% ,(ik unlee .1 FATED
- THERM AL PoWEP.( 0t64 I Chk 4 6.ul ab' , (Xd 2, (gc)g 3 ad dhd
i 2.1 SAFETY LINITS
\
ansEs 2.1.1 REACTOR CORE i The restrictions of this Safety Limit prevent overheating of the fuel and ' possible cladding perforation which would result in the release of fission ' products to the reactor coolant. Overheating of the fuel cladding is prevented my restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface tamperature is slightly above the coolant saturation temperature. Operation above the oper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DN8) and the resultant sharp reduction in heat transfer coefficient. DNS is not a directly esasurable parameter during operation and therefore THERNAL POWER and Reactor Coolant Tcaperature and Pressure have been related to DNS,tt7: qh tM i 5 1 n rr: M^ %;.. 7.% C 1 Z ;;77;ht%7. in %is Malp* ' has been developed to predict the ON8 flux and the location of DNB for axially unifors.and nonunifore heat flux distributions. The local DN8 heat flux ratio (DNBR) is defined as the ratio of the beat flux that uould cause DN8 at a particular core location to the lecal heat flux, and is indicative of the
" "II" **
- for eFAhl a4 %c wit 9 2 corre,lakien Ihr MAN 4k 5 [us!
' . The DNB desigdasis is as follows: there sust be at least a 55 percel.t probabi11ty that the etnte R of the limiting rod during Condition I and II C events is' greater than or to the DNSR iisit of the DNB correlation being used (the WRB-1 correlatio this application). The correlation DN8R 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 sinimum DN8R is at tho - v 0N8R limitg(j.lSr ben +c WRD-lmed 40R3-2 c,,,e.kgn comaahsy, In meeting this design basis, uncertainties in plant operating parameters, .
' nuclear an~d thermal parameters, and fuel fabrication parameters are considered '
statistically such that there'is at least a 95 confidence that the minimum DN8R - for the limiting rods is greater than or equal to the DN8R limit. The uncer-tainties in the above plant parameters are used to determine the plant DN8R uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DN8R value which must be met in plant safetv analysis using values of input parameters without uncertainties. LN5ER/t I) The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant systen pressure and average temperature for which the minimum design DN8R is no less than 1.M 7; : t;,- k;l n il n 1. % f r : W e h n " ,0 cclo.g'
- magosk,or the average enthalpy at the vessel axit is :;r' t the enthalpy of saturated liquid. leH M AnAel oD's #
allowanc.5 e is included u N for 8I'r an Y tr$a*".'"necease in Fh AH'at'I reduced IO pow
. expression:
y -
.p . . mf 6.a ' w..x l
BRAIDWOOD - UNITS 1 & 2 8 2-1 L l
INSERT 1 The design DN6R values are 1.34 and 1.32 for a typical cell and a thimble cell, respectively for 0FA fuel, and 1.33 for a typical cell and 1.32 for a ! thimble cell for the VANTAGE 5 fuel. In addition, sargin has been maintained in both designs by meeting safety analysis DNBR limits of 1.49 for a typical I cell and 1.47 for a thimble cell for 0FA fuel, and 1.67 and 1.65 for a typical cell and a thimble cell, respectively for the VANTAGE 5 fuel in perfoming : safety analyses. ' I l l s
, , , , ,~,,--,-.,c..,-,.,--. - - - - - , - - - . - - , - - - - - - - - -- -- --- - - - - - - - - - - - - - - - - ' - ' - - ^ - - -
SAFETY LIMITS g SAsts- 'i REACTOR CORE (Continued) ANf le fs bei.3r $hMA F fuel Where 1s the fraction of 4ATED THERMAL POWER. l 1 These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assains the axial power fabalance is within the limits of the : f (AI) function of the Overtemperature trip. When the axial power imbalance < is not within the tolerance, the axial power imbalance effect on the Overten- ' perature AT trips will reduce the Setpoints to provide protec, tion consistent with core Safety Limits.
- l-2.1.2 REACTOR' COOLANT SYSTEM PRTS$URE The restriction of this $afety Limit protects the integrity of the .
o Reactor Coolant System (RCS) free overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant free reaching the
- l containneht steosphere, t
The reactor vessel, pressuriser and the RCS piping, valves and fittings aredesignedto'$ectionIIIoftheAIMECodeforNuclearPowerPlantswhich .. permits a maximum transient pressure of 1105 (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements. The entire RC$ is hydrotested at 3110 psig,125K of design pressure, to demonstrate integrity prior to initial operation, s l g SRAIDWD00 - UNITS 1 & 2 8 2-2
[ , l , REACT 7VITY CONTROL SYSTEMS ' 190fitaTOR TtMPitAtutt cotFFICIENT
- i. LIMITING CONDIT10N FOR OptRATION b <
3.1.1.3 The moderator temperature coefficient (MTC) shall be: ' e. 4Less
.,.iposhive than 0 &VU'F for the all rods withdrawn M ;i- '.;
i;.^. G;e het zero THERMAL POWER condition, o,r !
- b. '
Less negative than -4.1 a 10 4 & V W 'F for the all rods withdrawn, 4 end of cycle ?ife (E0L), RATED THERMRt POWER condition. APPLICABILITY: 5pecification 3.1.1.3a. - MODES 3 and 2onlyW. Specification 3.1.1.3b. - ICOES 1, 2, and 3 only#. , ACTION:
- a. With the M7C acre positive than the limit of Specification 3.1.1.3a.
above, operation in M00E51 and 2 mayfproceed provided: '
- 1. '
Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than 0 A V U'F , i' ~ within 24 hours or be in NOT STANDBY within the.next 6 hours. These limits ofwithdrawal Specification limits3.1.3.6; shall be in addition to the insertion 2. The control rods are maintained within the withdrawat limits established above until a subsequent calculation verifies that , the MTC has whhdrawn been restored condition; and to within its limit for the all roes
- 3. A special aeport is prepared and submitted to the Coseission !
pursuant to Specification 6.9.2 within 10 days, describing the - value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for : restoring the positive MTC to within its limit for the all rods s withdrawn condition. r___ i -
\' b. With the MTC more negative then the limit of Specification 3.1.1.3b.
above, be in HDT SHUTDOWN within 12 hours. ' 4, %c, prev't515W d Sftdt(i(Mdre 3.c 4 Wre, het aeplicable.. e l "With K,gg greater than or equal to 1.
#5ee Special Test Exceptions Specification 3.10.3.
BRAIDWOOD - UNITS 1 & 2 3/4 1-4
. _ - . _ _ _ . . _ _ _ _ _ . _ . _ _ _ . _ _ _ . - _ _ = _
.4
(. - - REACTIVITY CONTROL SYSTEMS 4
- o *
$URVEILLANCE REQUIREMENTS 4.1.1. 3
###
- The MTC shall be determined to be within its limits during each fuel
' i a.
cN8tc4ed MTC. ioedadib AdmidisOS The MTC shall be measured and coa. pared to the 80C"h11mit
,_ Specification 3.1.2.3a. , abovej prior to initial operation above 55 of RATED THERMAL POWER, after,each fuel loading, and b.
The NTC shall be measured at any THERMAL POWER and compared to
. -3.2 x 10-8 A V U'F (all r !
tion) within 7 EFPD after aching withdrawn RATED THERMAL POWER condi-tration of 300 ppm. an eq,uilibrium boron concen- , is more negative than -3.2 In the event this comparison indicat.es the MTC and compared to the EDL MTC 10-* AUU'F, the MTC shall be remessured, ,
~
least once per 14 EFPD duringithe remainder of the fuel cy
~ Tin, )l*.IY Cm lo'ft kg'3mg,f8 :P C%
b 4'af 3 , c l u L' ; , t s l L1 l 1. i i 9 BRAIDWOOD - UNITS 1 & 2 3/4 1-5 _-+___m._._w_ _ _ _ _ _ . _ _ _ _ _ . . , _ _ - - _ . . , , . - - , , , . , , - - - . . , . . ,,y , , , , . , _ , , , , , , _ - . . . . , , , , , , _ . . - . , , - .
l P REACTIVITY CONTROL SYSTDt$- , (- , ; 800 DROP IIME ;
' LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length shutdown and control rod drop time from i
the fully withdrawn position shall be less than or oeval to 2.4 seconds fres beginning er accay of stationary gripper coi r voltage to dashpot entry swith: ,
- a. 7,,, p a w m n or q ual u M , a d
- b. All reactor coolant pumps operating.
APPLICABILITY: MDDES 1 and 2. ACTION: e. With the rod 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 MDDE 1 or 2. l
. b. With the rod drop time within limits but determined with three >
reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 66% of RATED THERMAL POWER. l -; .y
$URVEILLANCE REQUIREMENTS ,
4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality: a. For all rods following each removal of the reactor vessel head,
.b. For specifically offeeted individual rods following any maintenance l on or modification to the Control Rod Drive Systas which could affect the drop time of those specific rods, and
- c. At least once per la sonths.
{ k ( Q h'd l b$ II A ON E- 0- 2,) IfC0 $ (Oh i C if=* 3 & a Mer ) 0M 2. C.gde. 2. emd akr)
- 1 l
l-l 1 SRAIDWOOD - UNITS 1 & 2 3/4 1-19
l l , gl , 1/4. 2. t W.AT PLW W T O w etEL FACTOR - F;ft) LIMITI E CONDITION FOR OPERATION j
'3.2.2. Fg(I) thall be Itained by the following relattenships Fq (!) 5 ) (E(!)) for P > 0.5, and ( Od 1 cyt,1t Z ,
Fg (!) i [4.64) (K(Z)) for P i 0.5. a n d O n't4 7 Ci$Eh Where: pe EE EI
- 1 RATED MlUML POWER !
and E(Z) is the function obtained free Figure 3.2-2 for a given ; core height location. 1 . . , [ APPLICABILITY: ICDE 1. Eut!!: .
.With F g(Z) exceeding its limit:
, .a. Reduce THEIDEL POWER at least N for each NgF (Z) exceeds the limit I
within 15 einstes and sieflarly reduce the Power Range Neutron-Flux-High Trip-5etpoints within the next 4 hours; P0nIR OPERATION -
. 't - say proceed for up to a total of 72 hours; subsequent POWEk j OPERATION any proceed provided the Ovemower AT Trip setpoints have been reduced at least N for each 3 Fg (Z) exceeds the limit; y .nd ,
- b. Identify and correct the cause of the out-of-liett condition prior to increasing THERMAL' POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit. g Fe(1)1[2 5OD(G kr P >0.5, owd ( Un'd I C F' I 3 8h5 P
' e'5* b U " N 2 C d$ t 2 a d Fe Cl) *[S 803E Kb' gg L .
1 .A BRAIDWOOD - UNITS 1 & 2 3/4 2-4 L l>
w '
'e ll n,
.. i I
(. f L"Eik' ":- Elf M FFf_ hf,P E-$ [~ :f M; E.7 Ej H l *:" E , sr Y e r,,5 p- =y !? 1 [ Q E- N = 7 W= ? . :5 is ;. t c s__ .,_ .3 y = . . . _ _ . -
- m. e. :: _1 s = =,_ . . -. - ,o._ - - - <-
; 7 ve
. v.p. . ,. m..,
c q . ;- . . +. I i s t g [ ==-. p [ -_[ i-=. 5g 9 ( i f h .: j;; E
-. =
{'
= . .. - - - - - _ - ' = .. , - . ,
h 5
^ ^'
x{ C 5 k 3. . . . rJd k b 3 2 A ge it= E_ ( i i ._ . m
.n m g.(,=g7
--m_.
P;i=11
===. . -
g_m, , (_. gi m. 7- r. . , y : -
, s P'@'t Eb i i N eME"=c "e M SM t m- t '
1 2Cf./2fBiiBainM s l'MzMkn=r s Eiki + :o - gfjwpi
===,e w,ee;u e - = u s a rw 21p h n= : 4 x e e- + e=:i s_ .: m .t. m 4 -
Pai-i s~misl E E PE~ E ~ E- - i sii~nEicma .. ini r . Er E F1 F =, g g
=" m M M E s ?>*>P ? % *^I W P M t' h i "
.i*
l l #M.bBd-D'b E r- E'W M M D'ME?>o- d.I
,g ~
EREBB R:.idd=n= M bi:$ldt= n 5'Ett ; E c 59M55 fuk. mmh-irhiMkic5"s=Mi=.'E i'ld +
- c ,
t LEEEEE'!!E Fi L*5EElE& M M'5"RLiMtaTRWhirK es eE E E M'e'i'E"i E t M # c W
*E8 .
o BAiB 2"Lhb E+:=TFS E2 E={gg{ [ Egg ( = g l E d t 5 2 5 P M .I d @ s 4 L i @ l' 4 4- N & - i h' , 3 II E 5 Ersiix M? 6. sere ntith F2 hrw F,g>.> m BB ECEL"lNJ""'!"""'71Es%T' Tr, iu. 5555 EMP75PI'l""NFE19E""#FF" "; a* , EMBE WEE"42!WPplWlPJERFMTEW ""4"M i g < . Uc_ = . = > = EEB.i'. d APA >M ' s = m = = _. w , g E
'L'"EEE g Eg$kE$$bN52 R EE"'
R B E =- Gi.li,t:A.E.121 D. M.g _$ $ $ b b N a s m a n e c h t- s e a m s mommsw. MBBB - EC'5""$E5bME r m = = =_ = iM116.EiMW y =====m w E = c Fe'aEr=lE3simE%EEEaH iiaF a-r;EiGEEnsfu_._ EM u-s saEg.P mm:ssui W'5Fer Eac c;- =c=.c-ar=c=asesesseses c=c = s:.s um -as_ms:xc-- - seers sssa == ' EEM5 EE""EEE""=WE"ER*"s"=LE""O"""Er=' a . w .e.
=
=
w
=,
. e
- tr@d onznynvoN
- 12W t
su mueeD - UN ITS 1 & 2. 3/4 2- f
l_ POWER DISTRIBUTfD. ;1MITS
' SURVEILLANCE REQUIREN'.NTS (Continued) y )
- 2) When the F is-less than or equal to the F RTP limit for the appropriate measured core plane, additional power distribution '
anos shell be. taken and F compared to F RTP and 7 at least ! once per 31'EFPD.
- e. The F,y limits for RATED THERMAL POWER (F,RTP) shall be within the !
.. Wit 4providedintheOPERATING1,1MITSRE50RTforallcorepienes 5 containing Bank "O" control rods and for all unrodded core planes; u f. The F,y limits'of Specification 4.2.2.2e. above, are not applicable f, in the following core planes regions as measured in percent of core height from the bottom of the fuel:
i
- 1) Lower core region from 0 to 155, inclusive,
- 2) *
, Uppercoreregionfrom85to1005, esced VAMAg 3 assembly.inclusive'lFM(itdal
- 3) Within t 25 of gri(d plane regions a such that no more'than 20% of' the total core height in the center core region is affected, and
- 4) Core plane regions within i 25 of core height (* 2.88 inches). '
about the bank demand position of the Bank "D" control retis,
- g. With F exceeding F , the effects of F on Fg (2) shall be. -
evaluated to determine ifgF (Z) is within its limits. 4.2.2.3 When F determinations, an overall measured Fshall (2) be g(Z) is measured 9 fordistribution other than 9 obtained from a power map Fand increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. *
/
A 3 y g , 1 g ka ) . b i. % 1 % ~ 1 BRAIDWOOD - UNITS 1 *, 2 3/4 2-7 Amendment No. 15
i l
. i POWER DISTRIBUTION LIMITS l
3/4.2.3' RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT a LIMITING CONDITION FOR OPERATION 3.2.3 ' Indicated Reactor Coolant System maintained as. follows for four loop opera (tion.RCS) total flow rate and Fh shall - [ a .RC$ Total Flowrate 1 390,400 gpe, and b b. F L < 1.55 (1.0 + 0.3 (1.0-P)] O r C M b
~
f,6g Cl.o + 0 3 (l.o-PD Or VAN F beI l Measured values of Fh are obtained by using the novable incore s L detectors. An appropriate uncertainty of 42 (nominal) or greater . shallthenbeappliedtothemeasuredvalueofFhbeforeitis compared to-the requirements, and , i I THERMA. P)WER p = RATED TNE:IMA. POWER L ,
- APPLICABILITY: MDDE 1.
1 ACTION: l With RCS total flow rate or Fh outside the region of acceptable operation:
- a. Within 2 hours either:
- 1. Restore RCS total flow rate and F or to within the above limits, i 2.
Reduce THERMAL POWER to less than 505 of RATED THERMAL POWER and rectuce the Power Range Neutron Flux-High Trip Setpoint to less than or e , next 4 hours. qual to 55% of RATED THERMAL POWER within the : t ' e t 8 BRAIDWOOD - UNITS 1 & 2 3/4 2-8
p , i RDCTIVITYCQNTROLSYSTIMS sAsts ' DODERATOR TEMPERATURE COEFFICIENT (Continued) . The most negativa NfC value equivalent to the most positive moderator o density coefficient ( WC), was obtained by incrementally correcting the MC used'in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MC associated with a core condition of all rods inserted (most positive WC) to'an all rods withdrawn- , condition and, a conversion for the rate of change of moderator density with
. temperature at RATED THEWEL POWER conditions.. Th'-'value of the M C was then transformed int.o the limiting NTC value -4.1 x 10
- Ak/ar-i. The MTC
- value of -3.2 x 10-4 Ak/k/*F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppe equilibrium boron -
concentration and is obtained by making these corrections to the limiting MTC value of -4.1 x 10-* ak/k/*F. f men _3 ' ' 5 M==illance Requirements for esasurement of the F M tb Lemmng ! g and near within its the=Jend
" of i the
. m rum - S = ~' -me to tonfits that the NfC remains s coefficient cha.g.6 .im %
- j. , n RCS boron concentration associated with i.e1 burnrincipallytothe{
I '
-3/4.1.1'4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical 'I with the Reactor Coolant Systes average temperature less than 550*F. This '
l limitation is required to ensure: (1) the moderator temperature coefficient is ' within its analyzed temperature range. (2) the trip instrumentation is within i its normal operating range (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, (4) the reactor vessel is above its ' minimum RTgy temperature, and (5) the plant is above the cooldown steam dump permissive, P-12. ' L L 3/4.1.2 30 RATION SYSTEMS
. ?
The Boron Injection Systes ensures that negative reactivity control is i available during each M00E of facility operation. The components required to perfore this function. include: (1) borated water sources, (2) charging pumps (3) separate flow paths (4) boric acid transfer pumps, and (5) an emergency, power supply from 0PERABLE diesel generators. With the RCS average temperature above 350*F, a minimum of two *>oron injection flow paths are required to ensure single functional capabilit the event are assumed failure renders one of the flow paths inoperable. yThe in boration capability of either flow path is sufficient to provide a SHUTOOWN MARGIN from expected operating conditiont of 1.3% Ak/k after menon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 15,780 gallons of 7000 ppe borated water from the boric acid storage tanks or 70,450 gallons of 2000-ppe borated water from the refueling water storage tank. BRAIDWOOD - UNITS 1 & 2 8 3/4 1-2 4
p aw- - x - y) e 'e i m-+.46 A- .a ~ & '
-2sMa- A,---
l l
)
1 INSERT 3 1
.c
~~
The surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confim that the MTC can be maintained within its 11mics. The 80L MTC measurement combined with the predicted MTC with core burnup can be used to impose administrative limits on rod withdrawal to ensure that MTC will always be~1ess positive than 0AX/X/*F. This coefficient charges slowly due principally to the reduction in RC5 boron concentration associated with fuel burnup. , a
}
l-1' 1 t
t 3/4.2 POWER BISTRIBUf!M LDGTS I RAsts
' The specifications of this section provide assurance of fuel integrity !
during Conditten I (Notesl Operetten) and II Incidents of liederate Frequency) . events ty: equal to ".i{1) 7;. ; ^, A . esistatning tk minima
". rf " . : 7; enesen(, DISR e'r t x?? during nores) in the core gr' sporetten fuel pelle la short-ters transients, reture, and cleading and (2)l properties to within assmedIta eschanica design eriter a. n additten,11ef tfas the esak linear power density during Conditten I eventa revides assurance that the initial sendittens assmed for j the LOCA analyses are est and the ICCS assaptance criteria limit of 2200*F is !
est exceeded, the ettprepet' ale, DN8R 1hnM 6
- Nh LQ 1 The definitions of certain het channel and peaking facters as used in
! these specificattens are as fe11eus.: 1 FO(2) Neat Flum Het Channel Facter, is defined as the maximum local ' heat flux en the surface of a fuel red at core elevatten 2 divided by the average fuel red heat flux, elleving for annefacturing . l talerances en fuel pellets and reds; F
'[ Nuclear Enthalpy Riss Not Channel Factor, is defined as the estio of the integral of linear power along the red with the h10 hest integrated ,
power to the average red power; and .. . i (.. .. F%(2) Radial to average peaking Facter power is defined density as the ratio in the herimental of peak plane power at core density elevation 6. 2/4.2.1 AI! L FLM DIFFre w s The liep pn ARIAL FLUX DIFFERENCE (AFD) assure that the F g (Z) apper bound : L envelope of times the meres 11aed axial peaking facter is not exceeded during either normal operation er in the event of menon redistribution following Power changes. ' Target flux difference is determined at equilibrium senen sendittens. N full-length rods any be posittened within the tore in accordance with their respective insertion limits and should be inserted near tleir neraal position for statestate speratten at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THEIDEL POWER is the target f1ur difference at RATED THEN14L POWER for the assectated core burnup eenditions. Target flux differences for other THERMAL
' p0WER levels are obtained by multiplying the RATED TIElpl4L p0WER value by j the appropriate fractional THEIDEL POWER 1evel. The periodic updating of ,
the target flux difference value is necessary to reflect core burnup l h censiderations. 1 . l l g SRAIDWOOD = UNITS 1 & 2 8 3/4 2-1 L 1
.o POWER DISTRIBUTIgil LIMIIS RAtt$ * '
NEAT PLlX si CHA10lf. FACTOR. and RCS FLOW RATE Am 1821*** fiiTriAlpY t!st nm ensu.tamm wenunues) i
- c. i The conteel red insertion lietts of Specification 3.1.3.5 are asintained,and i
- d. .\
The axial power distribution, expressed in teras of AX1AL' FLUX ' O!FFERENCE, is maintained within the limits. F"g will be asintained within its 11sita provided the Conditi
- d. above are esistained.m The combinetten of the GCS flew requirement (39 and theest.
will.be requirementgen F" guarantee that the DER used in the safety sis analy , fell red how son:penalty is not applied to the final value of F" g - for the Fuel red the methods described s reduce the value of the 054. h6 predictions with ' 1979 for the 17 17 Optimited 8691 Revisten 1 " Fuel w Evaluation,' July emblies
- WERT reduction en'0$ R will be less than that the fuel red how i At higher..burnups, the decrease in IldD/NTU assembly average burnup.
t na ; [- fission product inventary as ceaponsate fa s and the buildup of bow reduction in DNBR. Y There is a egin available between the 1.32 and 1. ' and the 1. der 1.49 safety analysis ONBR limit. Use of the R fue D E bow R limits. analysis limits.n reduction.still leaves a 5 margin in DER between design lie , i The.RCS flev requirement is based on the loop sinieua esasured flow rate of g7,800 sps which is used in the taproved Theme) Design procedure described l in FSAR 4.4.1 and 15.0.3. A precision heat balance is and is used.te calibrate the RCS flow rate indicators. perfomed once each cycle feedwater venturi which eight not be detected seul potential' fooling of the precision heat baIance in a non-conservative ma,nner.d bias the results- from the 5.15 is assessed for potential feedwater venturi fouling.Therefore, a penalty of r uncertainty of 2.5 has been included.in the 1eep etnism asasured flew rate toA account for potential undetected feeduster venturi feeling and the use of the RCS flow indicatars for flew rete verification. Ag fou11ag which might bias , the RCS f1sw thte asesurement greater than 0.35 can be detected by annitorin and trending serious plant performance parameters. If detected action shal' be teken, before performing subsequent precision heat balance measu,rements, i.e either the effect of fouling shall be quantified and compensated for in the RCS flow rate esasurement, or tm venturi shall be cleaned to eliminata the fauling possible flow reductions ene to av rapid core cred buildup. S tation shall be calibrated within seven days prior to the perf calorimetric flow esasurement. This requirement is due to the fact that the ' drift effects of this instrumentatten are not included in the flow measu uncertainty analysis. This requirement does not apply for the instrumentation whose drift effects have been included in the uncertainty analysis. BRAIDWOOD - 44 TITS 14 3 8 3/4 2-4 " .
-~w-w 6-- e s w, m m-s- v e .vv,,_,,m-..-.--,-,-,_a _ _ _ _ . _ _ _ _ _ _
A.
' .i INSERT 2 1
Margin between the safety analysis limit DNBAs (1.49 and 1.47 for the OFA fuel typical and thiebte cells, respectively and l.67 and 1.65 for the VANTA615 typical and thieble cells) erJ the design limit DNBRs'(1.34 and 1.32 for the OFA fuel typical and thlable cells, and 133 and 132 for the VANTAGE 5 . . 1 fuel typical and thimble cells respectively) is maintained. A fraction of this margin is utilized to accemodate the transitten core DNBR penalty (maniana of 12.95) and the appropriate fuel red bow DNSR penalty ; (less than 1.5% per WCAP 8691. Revision 1). The rest of the margin between design and safety analysis DNSR limits can be used for plant design flexibility. J : 1 i I l
. . . . _. __._-,. - -- - ~ . - - - - - - - - - ~ - - - - - ~ ~ - - - - - " ~ ~ ~ ~ ~ ' ~~ ~~~ ' '
1
+
lj i POWER O!STRIBUTION LIMITS Basts ! t NEAT F.UX @T CNANNH. FACTOR. and RCS FLOW RATE " NUCt8AB ENTHALpY RISE noT cnuusE. FEIos gentinued) '
.5esh n 3.1.3 The limitsof 4,46 for Fg# oet d not assume any specific uncertainty on the measured value of F#g. An appropriate uncertainty of 45 (nominal) or greater la added to the asasured value of F" g before it is compared with the requirement. ;
When an 0F asasurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. {
- An allowance of 85 is appropriate fw a full-core aP54 taken with the incere Detector Flux Mepping Systas, and 4 i 35 allowance is appropriate for manufacturing tolerance. <
The Radial Peaking Factor, F, (!) is measured periodically to provide j
- assurance that the Not Channel FZ (y) remains within its limit. The F limit i for RATED THERMAL POWER (F TP)qasprovidedinSpecification3.2.2waf !
determined free expected power control maneuvers over the full range of burnup i conditions in the core. The 12-hour periodic smveillance of indicated RCS flow is sufficient to i {
< Iimit, flow degradation which could lead to operation outside the acceptable detect i
3/4.2.4 00ADRANTPOWERTILTRAT19, I The QUADRANT POWER TILT RAT 10 limit assures that the radial power dis-tribution satisfies the design values used in the power capability analysis, i Radial power distribution esasurements are made during startup testing and periodically during Power operation. The limit of 1.02, at which corrective action is required and linear heat generation rate protection with x y plane power, provides tilts. A DNS limit of 1.02 nas selected to provide an allowance for the uncertainty associated with the indicated power tilt. The 2-hove tise allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tien of a dropped or sisaligned cot: trol rod. In the event such action does not correct the tilt, the margin for uncertainty on Fq si reinstated by reduc-ing the maximum allowed power by 35 for each percent of tilt in excess of 1. . For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to ct,nfirm that the normalized synestric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitorin flux nap or two sets of four symmetric thimbles.g The is done with of two sets a full fourincere syn-metric thimbles is a unique set of eight detector locations. Thege locations are C-8 E-5. E-11, N-3, N-13, L-5, L-11, h-4. EmmW.%x%5RnJw m
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Attpchment 3
'NON-LOLA ACCIDENT ANALYSIS- - . M THE BYRON /8AAION000 $TATIONS UN!TS 1.AND 2 TRANSITION TO NESTINGNOUSE 17x17 VANTAGE'5 FUEL ASSEM J
j
'l i- j l
i f% s 07211*4.A00717 e s e,~ . - - - , ~ - ----_..e--- - - - . - - -- - - _--- --,- - . . - - - - - --- - ---- - ,-- - - . - - - - - .--,+ w
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I.AE 1 2 CONTENT 5 _
\:
Section Descriotion i Pace !
-15.0- ;
ACCIDENT ANALYSES AND EVALUATIONS I 15.1- NON-LOCA OVERVIEW '
] 15.1.1 Introduction and Summary
'15.1.2 15.1-1
.0perating Parameters '
15.1.3' 15.1-3 q References ' 15.1-3 15.2 NON-LOCA ACCIDENT EVALUATION . 15.2.1- DN8 Events e ' 1' 5.2.2 - 15.2-1
'Long-Ters Heat Removal Events 15.2.3 15.2-2 Dilution Events
'15.2.4- 15.2-2 >
Steamline Break Events 15.2.5 15.2-3 References * \ '15.2-4 i 15.3 DECREASE IN REACTOR COOLANT SYSTEN FLOW RATE L 15.3.1 l Partial Loss of Forced Reactor Coolant Flow 15.3-1 4 15.3.1.1 Identification of Causes and Accident Description 15.3-1 L 15.3.1.2 Analysis of Effects and Conscouences ! 15.3.1.3 15.3-2 Conclusions 15.3-4 15.3.2 Complete toss of Forced Reactor Coolant Flow . 15.3.2.1 15.3-4 Identification of Causes and Accident Description 15.3-4 15.3.2.2 Analysis of Effects and Consequences 15.3.2.3 15.3-6 Conclusions 15.3-7 h 1ss0v:10/Os044:
1 1 l g , ! \ r ' TABLEOFCONTENTS(Cont) l L l Section Description ) Paes 1 j 15.3.3 . Reactor Coolant Puep Shaft Creek 1 ,' 15.3.3.1 15.3-7
)
# Identification of Causes and Accident Description 15.3-7, '
/ 1 15.3.3.2 Analysis of Effects and Consequences b 15.3.3.3 15.3-8 ; Conclusions L 15.3-12
, 15.3.4 :
Locked. Rotor.with Loss of Offsite Power 15.3-13 ? 15.3.5. References '
, 15.3-13 ,
L 15.4 SPECTRUM 0F'R00 CLUSTER CONTROL ASSEMBLY L. ; EJECTION ACCIDENTS . 15.4.1-1!dentificationofCausesandAccidentDescription 15.4-1 1
.15.4.1.1 Design Precautions and Protection {
L 15.4.1.2 15.4-1 Lieiting Criteria. 15.4.2 15.4-3 AnalysisofNfectsandConsequences 15.4.2.1 15.4-4 Calculation of Basic Parameters 15.4-7 15.4.2.2 Results 15.4.3 15.4-10 i Conclusions L 15.4.4 15.4-13 i I. References ' 15.4-13 1 i t :- , ; t l L l i- / ll It:0v:10/050449
TABLEOFCONTENTS(Cont) { i i Table l Description I 15.3-1 Time Sec'vence of Events for Incidents Which Result in . a Decrease in Reactor Coolant System Flow
'15.3-2 Summary of Results for Shaft Break Transient 4 -
q 15.4-1: Time Sequence of Events for the Spectrum of RCCA-Ejection Events 15.4-2 Parameters Used in the Analysis of the Rod Cluster Control Assembly Ejection Accident L i
)'
i r 4 s P G i
.R' 1:40r.10/0:03:s
q a , 14 . TABLE OF CONTENTS (Cont) ! Fiaure Description
~
c 15.0-1 Illustration of Overtemperature and Overpower. Delta T Protection l 15.3-1~ Flow Transients for Four Loops in Operation, Two Pumps Coasti! 15.3-2
- Nuclear Power and Pressurizer Pressure Transients for Four Loops )
Operation. Two Pumps Coasting Down 15.3-3 Average and Hot Channel. Heat Flux Transients for Four Loops in t Operation, Two Pumps Coasting Down 15.3-4 DN8R Versus Time for Four Loops in Operation, Two Pumps Coas Down i 15.3-5 Core Flow Coastdown for Four Loops in Operation, Four Pumps Coasting Down l
, 15.3-6 Nuclear Power and Pressurizer Pressure Transients for Four Loops in Operation, Four Pumps Coasting Down ,
15.3-7 Average and Hot Channel Heat Flux Transients for Four Loops in Operation, Four Pumps Coasting Down ' [ L 3 15.3-8 DNBR Versus Time for Four Loops in Operation, Four Pumps Coastin ' Down 15.3-9 Flow Transients for Four Loops in Operation, Reactor Coolant Pump Shaft Break l i 1ssov:1D/0:03s0
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D, I _TA8LECF;. CONTENTS (Cont) ' i v Fioure
- Descriotion 15.3-10 Nuclear Power and Reactor Coolant Pressure for Four Loops in :
Operation, Reactor Coolant Pump Shaft Break
.t 15.3-11
. s ..
Average and Hot Channel Heat Flus Transients for Four Loops in Operation, Reactor Coolant Pump Shaft Break 15.3 , . Maximum Clad Temperature at Hot Spot for Four Loops in Operation, Reactor Coolant Puse Shaft Break ' 15.4-1 F Nuclear Power Transient, 80L HFP Rod Ejection Accident 15.4-2 Hot Spot Fuel and Clad Temperature Versus Time 80L HFP Rod. ' EjectionAccident 1 15.4-3 Nuclear Power Transient, E0L HIP Rod Ejection Accident
<. 15.4-4 Hot Spot Fuel and Clad Temperature Versus Time E0L NZP Rod ;
Ejection Accident ' L i I g l l I isaccio/osesse l _______:__=_______-__ . - _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ . _ . . . . _ . - . - . , - _ . . . - . . _ - . . - - .-
L 4 15.0 ACCIDENT " LYSES AND EVALUATIONS 15.1 NON-LOCA OVERVIEW
.15.1.1 Introduction and Summarv ,
' This chapter evaluates the effects of tho' complete transition of ;
Byron /Braidwood Stations Unita 1 and 2 from 17x17 0FA fuel to 17x17 VA , fuel on the UFSAR Chapter 15 non-LOCA accident analyses (Reference'1) a l subsequent reference analyus. ,All events analyzed and reported in the UFSA have been reviewed. This section presents an overview of the changes and
--4ffects of the fuel transition on the non-LOCA ' safety analyses. Section 15.2' contains~an evaluation addressing all the non-LOCA events not explicitly analyzed for the transition.
Section 15.3 presents the analyses of the loss '
'of flow events. Finally, Section 15.4 contains the analysis of the RCCA ejection event.
Increased Peakina Factors t / A Nproposed change from 1.55 to 1.65 for the Nuclear Hot Channel Factor 0 (F g) is evaluated as well as an increase in the Total Peaking Factor ' (Fg ) from 2.32 to 2.50. An increase in the full power F"g limit l does not directly affect the system transient response of.the non-LOCA , events. . Instead, the FN g limit is used in the determination'of the DN8R'for those events.in which DNB is the safety analysis criterion and is offset in the determination of the core thermal limits by the addition of DNS margin generated by the Intermediate Flow Nixing Grida (IFNs). The increase in F"g is not relevant for this non-DN8 related non-LOCA events. The L increase in F0 si explicitly considered in the analyses presented in Sections 15.3 and 15.4. AT Protection Setnoints g
- The current 0FA fuel core thermal limits are more restrictive than th 5 fuel core thermal limits.
( Therefore, the core thermal limits presented in E tenomia msasse 15.1-1
. i LI' 3
i I
~ the Teshnical Specifications (Figure 2.1-1) remain valid for the transition
'VMTAGE 5 fuel.
The current overtemperature AT (0T&T) and Overpour AT. I (0 PAT) reacter trip setpoints required by the Technical' Specifications wer
' con, firmed'to remain applicable for the introduction of VMTAGE 5 fue aforementioned peaking factor increases with the exception of the f(&I) penalty (see below). i As a result, none of the transients that rely on the AT reactor trips for reactor protection have been analyzed for the fuel transition. , Figure 15.0-1, provided for information bounds the safety l analyses assumptions and ' applies to all four unita during and after the transition to VANTAGE 5 fuel. This is discussed in greater detail in'Section )
15.2. . l-Asial Offsets As a result of new axial offset limits for the VANTAGE 5 fuel (incorporatin the peaking factor increases), the f(a!) setpoints presented in Section 2.2 ' L of the Byron /Braidwood Stations Technical Specifications were changed. Since , the f(AI) penalty is not assumed in any of the non-LOCA transients, there is ' no impact on the non-LOCA analyses. t Intermediate Flow Mixer Grid: i The introduction of the IFNs into the VANTAGE 5 fuel design increases the available DNS margin. Part of this Of8 margin was used to in'sure that the core thermal limits for VANTAGE 5 fuel with acceptable. g an F" of 1.65 are The other effect of IFNs is to increase the control rod drop time from 2.4 seconds to 2.7 seconds. The current non-LOCA safety analyses l presented in the UFSAR, with the exception of the rod ejection events, assume a conservative rod drop model wisich remains bounding even with the increase in the rod drop time as discussed above (2.4 seconds to 2.7 seconds). The new - E rod ejection analyses, presented in Section 15.4, assumed a 2.7 second rod drop tina. ' issor.to/esasse 15.1-2
L (- , ,
- t ,
m;. Reactivity' Limits The limits currently assumed in the non-LOCA eccident analyses for the important reactivity parameters are not expected to change as a result of the - fuel transition. The reactivity limits are verified on a cycle-by-cycle basis
-as part of the normal reload process (Reference 2).
I 15.1.2 Deeratino Parameters ' The following nominal operating conditions were assumed in the VANTAGE 5 transition: [
' CorePower(Nwt) 3411 System Pressure (psia) {
2250 Vessel Average Temperature'('F) i 588.4 Core Inlet Teeperature - non-ITDP ('F) 558.4
-Core Inlet Temperature - ITDP (*F) 559.2 Thermal Design Flow (gpe) 377600 Winimum.MeasuredFlow(gpe)~
" 390390 CoreAverageLinearPowerDensity(kw/ft) 5.43 Steam Generator Tube Plugging Level (5) 10 1
15.1.3 h i
- 1) "8yron/Braidwood Stations Updated Final Safety Analysis Report (UFSAR)," Docket Numbers 50-454, 455, 456, and 457, December 1988.
- 2) Davidson, S. L., et al., " Westinghouse Reload Safety Evaluation Wethodology " WCAP-9272-P-A (Proprietary) and WCAP-9273-A (Non-Proprietary), July 1985.
issov:to/osasse 15.1-3
\, .: -. . - . . - . - . - - - . . _ . - . . -
j j
)
15.2 Int-LOCA ACCIDENT EVALUATION ' I ( N fe11 ewing discussions identify these transients. net explicitly analys , for theanalyses previous VANTAGE 5 transition remain valid. and the be. sis for the conclusio ' 15.2.1 DNB Iyents !
' I N following Condition !! events are analyzed in the Byron /Braidweed Stati!
UFSAR(Reference 1)todemonstratethattheDNBdesignbasisissatisfied.As previously discussed, the core thermal limits rensin valid. N refore, the L CTOT and OPDT setpoint equations remain valid and no new analyses were l required for the fuel transition. The saisting analyses aise bound the i nominal operating conditions listed above and the increase in red drop time. {; i N eefore, the safety analysis DNS design basis is still met and the results and conclusions presented in the UFSAR for the following events remain valid. : Although the core thermal limits are not applicable to the Uncentrolled RCCA [ l tank Withdrawal from a Suberitical Candition and the RCCA Misalignment ! i ! (DroppedRod) transients,theDNSdesignbasiswasalsoconfirmedforthese two events. ! I-g U*SAR SECTION Feedwater System Malfunction: Reduction in Temperatur9 15.1.1 ' ! Feedwater System Malfunction: ! Increase in Feedwater Flow 15.1.2 . e Excessive Increase in Secondary Steam Flow 15.1.3 : Loss of External Load / Turbine Trip 15.t.3 Uncontrolled RCCA lank Withdrawal 15.4.1 From a Suberitical Condition t Uncontrolled RCCA lank Withdrawal 15.4.2 ' at Power RCCA Nisalignment ! 15.4.3 taeor.iomsines 15.t-1
i l M UFSAR SECTION Inadvertent Operation of the ECCS
- 15.5.1 :
l Inadvertent Opening of a 1 15.6.1 Pressuriser Safety or Relief Valve j i 15.2.2 Lone-Term Neat Removal Events i The long-tors heat removal events are aealysed to determine if the avatti feedwater (AFW) heat removal capacity is sufficient to insure that the peak i RCS pressure does not eaceed allswable limits and the core remains covered a! in a coolable geometry.
- , The increases ingF" andq F de not affect system transients, and thus, have no impact on these events. f The remaining effects of the ','ANTA$t 5 fuel transition will also have no discernable ispact j on these transients. Therefore, the results and conclusions presented in the l
UFSAR remain valid for the events listed below. ! i ((QLT pFSARSECTION Loss of Non-Emergency AC Power ! to Plant Availiaries 15.2.8 ! Loss.of Normal Foodwater 15.2.7
?
Feedwater Systes Pipe Braak . 15.2.8 I< i' t 15.2.3 Dilution Events
- i
}
l The dilution events are analyzed to demonstrate that the operators (or the l automatic mitigation circuitry) have sufficient time to respond prior to
- reactor criticality once an alarm is generated. The important parameters for
[ ' these events are the dilution flow rates, RCS peometry (volumes), boron worth. and shutdown margin. None of these assumptions will change as a result of tne { fuel tran'aition. Therefore, the increases in F g N :
, F g, and rod drop time will have no ispect.
Thus, the results and conclusions presented in the UFSAR for the events listed below remain valid. , l i P issorio/ososee 15.2-2 i
l g UFSAR SECTION Startup of an Inactive Reactor Coelant Pug 15.4.4 CVC3 Malfunction (Soren Dilution) 15.4.6 15.2.4 Steaaline treak Events N ro are four stenaline break-related analyses performed for the lyron and Braidwood Stations. , M t L l I UFSAR SECTION ; Inadvertent Caening of a SG Relief l i or Safety Valve 15.1.4 Steam System Piping Failure i (Double-Ended Rupture - Core Response) 15.1.5 i l Steamline Rupture Mass & Energy Releases i j Inside Containment 6.2.1.4 i' Steaaline Rupture Mass & Energy Releases Outside Containment for Reference 2 Equipment Environmental Qualification
- t N limiting core response analyses (UFSAR Sections 15.1.4 and 15.1.5) are ;
i performed at hot zero power conditions with the control rods fully inserted in the core and the most reactive red stuck out. Therefore, changes in rod drop'{ time are not applicable. An increase in the full power F" limit g k results in an increase of the stuck red peaking factor. N The effects of the : increased stuck rod Fg ave h been evaluated. The safety analysis { DNlt limit is met and the conclusions of the FSAR remain valid. i
' i For the mass and energy releases, the steam release calcu1ations are i
insensitive to the change in rod drop time and are primarily influenced by the core wide power level rather than local power variations within the core. j Therefore, the mass and energy events are not impacted by the increase in core peaking factors, and the results presented in the UFSAR and Reference 2 remain valid. i-i
)
tasov:ioassaas , 15.2-3
. l 15.2.5 References .
I i
- 1) *tyren/Braidweed Stations Updated Final Safety Analys.is Report (UFSAR),' Decket Iksbors 50-454, 455, 456, and 457, December 1908. -!
t t) Butler, J. C. and D. 5. Love, 'Stoseline 8ruk Mass / Energy Releases ) for Equipment Environmental Qualification, outside Containment,' I WCAP-10981-P (Proprietary) and WCAP-11184 (Non-Proprietary). Octob I 1905, i
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i L l l ! i v BYRoll/BRAIDWOOD STATIONS i FIGURE 15.0 1 ILLUSTRAT!0ll 0F OVERTEMPERATURE AND OVERPOWER DELTA T PROTECTION . { s.
...._m._. , - . _ _ _ . . _ . - - . . , - , _ _ _ _ , , . . , . - _ _ . - - - - - - . - - . _ - - . . . . , _ -
l t .
'15.3 MCMASE IN REACTOR @_MT SYSTDI FLOW RATE '
I i, A number of faults are postulated which could result in a decrease in reacto coolant system flow rate. i These events are discussed in this section. Oetailed analyses are presented for the most limiting of these events. { j i Discussions of the following flow decrease events are presented in Section 15.3: i I
- a. partial loss of forced reactor coolant flow, L b. couplete loss of forced reactor coolant flow, and
! i l'
- c. ructor coolant pump shaft break. !
i i Item a'. above is considered to be an ANS Condition !! e;ent, item b. an ANS ' Condition !!! event, a'nd item c. an ANS Condition IV event. 15.3.1 Partial Loss of Forced Reactor Coolant Flow ! i 15.3.1.1 Identification of causes and Accident Descriotion I A partial loss of coolant flow accident can result from a mechanical or electrical failure in a reactor coolant pump, or from a fault' in the power i supply to the pump or pumps supplied by a reactor coolant pump bus. If the reactor is at power at tb.e time of the accident, the immediate effect of loss of coolant flow is a rapid increase in the coolant tosperature This increase . could result in DNB with subsequent fuel damage if the reactor is not tripped promptly. , Normal power for two of the reactor coolant pumps is supplied through i individual buses connected to the generator, whereas the other two reactor - 4 tesotto/osess 15.3-1 !
.l e
coolant pumps are suppliind from offsite power. When a generator trip occurs, the buses which are nores11y fed from the generator are autoestically transferred to an offsite power supply. , The pumps will continue to supply coolant flow to the core. Following any turbine trip where there are no *I electrical faults or thrust bearing failure, which require tripping the generator from the network, the generator remains connected to the network for appreminately 30 seconds. [ The two reactor. coolant pumps normally fed free the j generator remain connected to the generator thus ensuring full flow for j appresinately 30 seconds after the reactor trip before any transfer is made. i o , This event is classified as an ANS Condition !! incident (an incidentj
. moderate frequency) as defined in Subsection 15.0.1 of the Byron /Braidwood j StationsUFSAR(Reference 1). #
The necessary protection against a partial loss of coolant flow accident is provided by the low primary coolant flow reactor trip signal which is actuated l in any reactor coolant loop by two out of three low flow signals. Above ' Permissive 4. low flow in any loop will actuate a reactor trip. Between approximately 105 power (Permissive 7) and the power level corresponding to l Permissive 4, low flow in any two loops will actuate a reactor trip. Above { j Permissive 7 two or more reacter coolant pump circuit breakers opening will i actuate the corresponding undervoltage relays. This results in a reactor trip i which perves as a backup to the low flow trip. . 15.3.1.2 Analysis of Effects and Con r m ee i i p MpthodofAnalysis : l One case has been analyzed: , j a. I Loss of two pumps with four loops in operation. This transient is analyzed by three digital computer codes. First, the LOFTRAN Code (Reference 2) is used to calculate the loop and core flow during L l L issswio/esasse 15.3-2 _ _ , _ _ - - . _ . _ _ _ _ . - . _ _ _ _ . _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _______________ _ ______ ____--___ .[
4 i
~
the transient. the time of reactor trip based on the calculated flows, the nuclear power transient, and the primary system pressure and temperature transients. j The FACTRAN Code (Reference 3) is then used to ca flux transient based on the nuclear power and flow from LOFTRAN.Finally, the THINC Code (Reference'4) is used to calculate the DNBA during the trans j based on the heat flus from FACTRAN and flow from The DNSR LOFTRAN. l transisnts presented represent the minimum of the typical or thimble cell. . i i This accident is analyzed with the improved thermal design procedure as { described in WCAP-8567. Plant characteristics and initial condition ! discussed in Subsection 15.0.3 of the Byron /Braidwood Stations UFSAR. Initial Conditions ' l i Initial reacter power, pressure, and RCS temperature are assumed to be at g i
' their nominal values, Uncertainties in initial conditions are included in the !
limit DNBR as described in WCAP-8567. ' l, Reactivity Coefficients ; I
/ i
-i A conservatively large absolute value of the Ogpler-only power coefficient is used (see Figure 15.0-3 of the Byron /Braidwood Stations UFSAR). This is
! l equivalent to a total integrated Doppler reactivity from 05 to 1005 power of 0.016 ok. ' { t The least negative moderator temperature coefficient (see Figure 15.0-2 of the . Byron /Braidwood Stations UFSAR) is assumed since this results in the maximum ! core power during the initial part of the transient when the minimum DN8R is I reached. Flow Coastdown ' ( The flow coastdown analysis is based on a momentum balahce around each reactor C coolant loop and across the reactor core.
- This momentum balance is combined
l with the continuity equation, a pump momentum balance and the pump ' characteristics and is based on high estimates of system pressure losses.
, issemiwosases 15.3-3
~ w - a --.-----,w s,,-,,,,- - - _ - - - , - - - - - - - - . - - - - - - , , , ~ - - - - - - - - - - - . - - - - ~ - - - - - - - - - - - . _ _ . - - - - - _ _ _ - - - - - - - - - -m - -
i
~
No single active
- failure will adversely effect the consequences-of the u accident. .
Results
) - Figures 15.3-1 through 15.3 4 show the transient response for the loss .k of reactor coolant pues with four loops in operation. The reactor is assumed to )
be tripped on a low flow signal, Figure 15.3-4 shows the DNBR to be alwa greater than the limit value. l' t For the case analyzed, since DNB does not occur, the ability of the primar coolant to remove heat from the fuel rod is not greatly reduced. f Thus, the ! average fuel and clad temperatures do net increase significantly above their respective initial values. f i " The calculated sequence of events tables for the case analyzed are shown on Table 15.3-1. The affected reactor coolant pue will continue to coast down, i and the core flow will reach a new equilibrium value corresponding to the number of pumps still in operation. With'the reacter tripped, a stable plant , condition will eventually be attained. Normal plant shutdown may than proceed. ! l 15.3.1.3 Conclusions e t l The analysis shows that the DNBA will not decrease below the limit value at - any time during the transient. Thus, no fuel or clad damage is predicted, and ; all applicable acceptance criteria are met. ! 15.3.2. Camolete Loss of Forced Reactor Coolant Flow + e 15.3.2.1 identification of Causes and Accident Descriotion k A complete loss of .'?cced reactor coolant flow any result from a simultaneous loss of electrical supplies to a.1 reactor coolant pumps. If the reactor is a 1 .at power at the time of the accident, the immediate effect of loss of coolant { 9 innovaamseses 15.3-4 l -_
\
,y .
l I 0 flow is a rapid incrosse in tho c. L*nt temperature. This increase could j" result in Die with subsequent fusi ,arge if the reactor wre not tripped promptly. 4 .. Normal power for two of the reactor coolant pumpa is supplied through
)
c'olant pumps are supplied free offsite power.froml o When a generator trip occurs, the buses wMeh are normally fed from the generator are automatically l
,. transferred to an offsite power supply. The pumps will continue to supij coolant flow to the core. Following any turbine trip where there are no electrical faults or thrust bearing failure which require tripping the generator from the network, the generator remains connected to the network{I appreaimately 30 seconds.
The two reactor coot tnt pumps norma)1y fed from the i generator remain connected to the generator thus ensuring full flow for 30 seconds after the reactor trip before any trcnsfer is made. ( t This event is clas.lfied as an ANS Condition !!! incident (an infrequent ! i i incident)asdefinedinSubsection15.0.1ofthaSyron/BraidwoodUFSAR.
'ho following signals provide the necessary protection against a complete lot of flow accident: i, i
- 4. i 4
Reactor coolant pump power supply underveltage or underfrequency. !
- b. l.ow reactor coolant loop flow. I t
i L ; The reactor trip on reactor coolant pump bus undervoltage is providea to f j protect against conditions which can cause a loss of voltaga to all reactor coolant pumps, i.e., station blackout. L This function is blocked below i E approximately 105 power (Permissive 7). l The reactor trip on reactor coolant pump underfrequency is provided to tris the reactor for an underfrequency condition, resulting from frequency ! disturbances on the power grid. { Reference 5 orovides analyses of grid ' frequency distu.bancer, and the resulting nuclear steam supply system (NSSS)
- protection requirements which are generally appilcable. ,
?
i m isvesess: 15.3-5 t
- 7. , .
pi .
\
g :: c 1 i 8 4 li' The reactor trip on low primary coolant loop-flow is provided to protect stainst less ef flow conditions which affect only one reactor coolant loop. 6 ' This function is generated by two out of three low flow signals per reactor coolant icg. 4
+ rip.
Above Permissive 8, low flow in any'leep will actuate a reactor Between approminately 105 power (Permissive 7) and the power level corresponding to Permissive 8, low flow in any two loops will actuate a ' reactor trip. 15.3.2.2 Analysis of Effects and Con r macu g' One case has been antlysed: s 1 l
- a. Loss of four pumps with four t w o in operation. 1 3
This transient is analysed by thro
- digital computer codes. First, the LOFTRAN Code (Reference 2) is'used to calculate the loop and core flow d; the transient, the tira of reactor trip based on.the caleclated flows, the i i nuclear' power transient, and the pr mary systes pressure and temperature transients. !
The FACTRAN Code (Reference 3) is then used to calculate! flua transient based on the nuclear power and flow from LOFTRAN. Finally, the THINC Code (Reference'4) is' used to calculate the DNBR during the transient based on the heat flus free FACTRAN and flow from LOFTP.AN. The ON ' transients presented represent the minimum of the typical or thimble cell. ! The method of analysis and the assumptions mais regarding dnitial operating 1 l corditions and reactivity coefficients are identical to those discussed in i l Subsection 15.3.1, except that following the loss of power supply to all pumps ! at power, a reactor trip is actuated by either reactor coolant pump power supply bus undervoltage or underfrequency. h k . Results
< i I
Figures 15.3-5 through 15.3-8 show the transient response for the loss of ' power to all reactor coolint pumps with four loops in operation. The reactor is assumed to be tripped on an a undervoltage signal. Figure 15.3-8 shows the DNBR to be always greater than the limit value. ' tesorio/ossses 15.3-6 "
1 i a' i j Q $, , For the case analyzed, since DNS does not occur, the ability of thej! coolant to remove heat from the fuel red is not greatly reduced. Thus, the e average fuel and clad temperatures do not increase significantly above their respective initial values. l l The calculated' sequence of events for the esse analyzed is shown on Tab 15.3-1. The reacter coolant pumps will continue to cast down, and nctural , circulation flow will eventeally be established. ; With the reactor tripped, a i stable plant condition would be attained. Normal plant shutdown may then* proceed. 15.3.2.3 Conclusions i The analysis performed has demonstrated that for the complete' toss of force! i reactor coolant flow, the DNBP. does not decrease below the limit value at an time during the transient. Thus, no fuel or clad damage is predicted, and all ( i applicable acewtance criteria are met. ' 15.3.3 Rosetor Coolant Puno Locked Rotor / Shaft Break - [ 15,3.3.1 Identification of Causes and Accident Description i' i The. accident postulated is an instantaneous failure of a reactor coolant pus! rotor or shaft. 3 Flow through the affected reactor coolant loop is rapidly reduced leading to an initiation of a reactor trip on a low flow signal. { Following initiation of the reactor trip, heat stored in the fuel rods , continues to be transferred to the coolant causing the coolant to expand, at I the s4me' time, heat transfer to the shell side of the steam generators is! , reduced, first because the reduced flow results in,a decreased tube side file > cuefficient and then because the reactor coolanc in the tubes cools the shell side temperature increasec (tuibine steam flow it reduced to zero f upon plant trip). The rapid expansion of the coolant in the reactor core, l combined with reduced heat transfer in the steam generators e,iuses an in w ge , l sesocio/oseass 15.3-7 l
. ~~ . - _ _ . _ _ . _ _ - - _ _ _ _ _ _ . _ _ . _ _ _ _ . _ _
i5
. c
)
& I N5 A' r t1 into the pressuriser and a pressure increase throughout the reactor cool system. ' ' The 'insurge into the pressuriser compresses the steam volume, *, k actuates the, automatic 6 pray system, opens the power operated relief jv and opens the pressuriser safety valves, in that sequence. The two power-operated relief valves are designed for reliable operatior, and would be{ espected to function properly during the actifient. { However, for conservatism, j their pressure reducing effect as well as the pressure reducing effect of ' spray is not included in the analysis. , This event is classified as an ANS Condition IV incident (a limiting fau! i defined in Subsection 15.0.1 of the Byron /Braidwood Stations UFSAR. : The conseguences of a locked rotor (i.e., an instantaneous seizure f a 'p , shaft) are very similar to those of a pump shaft break. The inittai rate of ; reduction of ' coolant flow is greater for the locked roter event. However, j
- witt, a broken shaft, the ispeller could conceivably be free to spin in the reverse direction. %e effect of reverse spinning is to decrease the !
.j steady-state core flow when compared to the locked retor scenario. Only one ;
analysis has been performed for this report and it represents the most l { L lisiting condition for the locked rotor and pump shaft break accidents. This j L analysis bounds both cases presented in the lyron/Braidwood Stations UFSAR (Section15.3.3and15.3.4). j L 15.3.3.2 . l \ - Analysis of Effects and Consesgn a g, l j Method of Analysis , i Two digital computer codes are used to analyze this transient. The LOFTRAN i Code (Reference 2) is used to calculate the resulting loop and core flow o transients following the pump shaft failure, the time of reactor trip based on ; the loop flow transients, the nuclear power following reactor trip, and to determine the ;4ak pressure. , The thermal behavior of the fuel located at the t core hot spot is investigated using the FACTRAN Code (Reference 3), which uses the core flow and the nuclear power calculated by LOFTRAN. The FACTRAN Code includes a film boiling heat transfer coefficient. iss w ivosanas 15.3-8
,_,n4ege-yw*"au- " * ' ' * ' _ . _ . _ _ _ , . - - - - - - --
_-.7
,a .
i Jt One case is an'a lyzed:
- a. .
Four loops operating, one locked rotor / shaft break. At the beginning of the postulated shaft break accident, i.e., at the time shaft in one of the reacte. coolant pues is assumed to fail, the plant is assumed to be in operation under the most adverse steady-state operatin condition, i.e., manieus guaranteed steady-state themal power, maximus ) j steady-state # pressure, and maximum steady-state coolant average temperatu
\
Plant sharacteristics and initial conditions are further discussed Subsection 15.0.3 of the Byron /Braidwood Stations UFSAR. { i For the peak pressure evaluation, the initial pressure is conservatively i estimatedas30psiaboveneainalpressure(2250 psia)toallowforerrorsin j the pressuriser pressure measurement and contrcl channels. ! Tnis is done to ootain the highest possible coolant pressure during the transient. The } pressure response shown in Figurs 15.3-10 is the response at the point in the { reacter coolant system having tne maximue pressure. i s For a conservative analysis of fuel rod thermal behavior, the hot spot l evaluation assumes that DNS occurs at the initiation of the trans( continues throughout the transient. This assuetion reduces heat transfer to the coolant and results in conservatively Mgh hot spot temperatures. { i No single active failure will adversely affect the consequences of the accident. , 1 Evaluation of the Pressure Transient t After pus failure, the neutron flux is rapidly reduced by control rod insertion . ! Rod motion is assumed to begin one second after the flow in the affected loop reaches 875 of naninal flow. No credit is taken for the pressure reducing effect of the presurizer relief valves, pressurizer spray. - steam duse or controlled feedwater flow after plant trip. Although these I operations are expected to occur and would result in a lower peak pressure, a ,n 1sser:nvessass o 15.3-9
, , - - w-.-,.--.., ,+..,-..---v-w-v,,-m <w.,- - . - - - , ..--,--..>~vm,m--.- -
additional degree of conservaties is provided by ignoring their effect. The pressuriser safety valves are full open at 2575 psia, ) i Evaluation of Hot Ssot Temerature in the Core Durine the Accident ! Although no rods'are predicted to be in DNb,' for conservatism for this accident, DNB is assumed to occur at the hot spot in the core, and theref! an evaluation of the consegcences with respect to fuel ros thermal' transients ! is r.orformed. Results obtained from analysis of this ' hot spot' condition I represent the upper limit with respect to clad temperature and airconium water reaction. i In the evaluation, the rod power at the hot spot is_ assumed to be ! 3.0 times tO average rod power (i.e., Fg = 3.0) at the initial core power level. ! s Film Boiline Coefficient
- y ]
t The file boiling' coefficient is calculated in the FACTRAN Code using the L Bishop Sar.dberg-Tong film boiling correlation. The fluid properties 'are evaluated at film temperature (aserage between wall and bulk temperatures). l ' The progran calculates the film coefficient at every time step based upon the i actual heat transfer conditions at the time. The r.eutron flux, system i pressure, bulk density and asse flow rate as a functica of time are used as ,! program input. L p For this analysis, the initial values of the pressure and the bulk density are . used throughaul the transient since they are the most conservative with respect to clad tosperature response. j
\
Fuel Clad Gao Coefficient The magnitude and time dependence of the heat transfer coefficient between fuel and clad (gap coefficient) has a prenounced influence on the thermal i results. The larger the value of the gap coefficient, the more heat is ' t isaov:to/esosse 15.3-10
d
.)
. .. i
\
transferred between pellet and clad. W- . Based on investigations on the effect of ! D the gap coefficient upon the maximum clad temperature during the transie the gap coefficient Oss assumed tn increase from a steady state value consistent with initial fuel tespera'ture to 10.000 8tu/hr-ft2 ..F at the I initiation of the transient. Thus, the large amount of er. orgy stored in the fuel because of the ama11 initia'. value is released to the cla initiation of the transient. Zirconium Steam Reaction
- The airconium-stoas reaction can beoose significant above 1800'F (clad -
temperature). k The Baker-Just parabolic rau equation shown below is used to ' define the rate of the airconi'un steam reaction. l- ! 2 2 ks33.3x10 0esp (- ( M D !
- where
- )
o ! w
- amount Zr reacted, og/ cat l t a time, sec 1 i,
I T = tem erature. *K I
, i The reaction heat is 1510 cal /ge. '
; i i
The effect of airconitam-steam reaction is included in the esicula1
" hot spot" clad temperature transient.
r l-i l
- l. !
+
l s tesov;tevosesse 15.3-11
- a. <
Shaft Break with Four L e e Q g Transtent results for this case are shown,in Figuren 15.3-9 through 1 The results of these calculations are aise summarized The in peak reactor coolant system pressure reached during the transient is that which would cause stresses to escoed the faulted cond limits. 2700'F. Also, the peak clad surface temperature is considerably less than It should be noted that the clad temperature was conse'.atively
-calculated assuming that DNS occurs at the initiation of the transient.
j The calculated esquence of events for i,he case analyzed is shown in Ta$ 15.3-1. ( Figure 15.3 9 shows that the core flow rapidly reaches a new equilibrium value. l With the reactor tripped, a stable plant condition will eventually be attained. Normal plant shutdown may then proceed. ')I s 15.3.3.3 Conclusigng, n
<a.
Since the peak reactor coolant system pressure re' ached during any l the transients is less-than that which would cause stresse ! ! the faulted condition stress limits, the integrity of the prima y coolant system is not endangered. j ! h. I Since the peak clad surface temperature calculated for the hot spot during the worst transient remains considerably less than 2700'F, the l core will remain in pla:e and intact with no loss of core cooling L capability. i c. Y The results of the transient analysis show that none of the fuel roes [ vill have DNBR's below the safety analysis limit value. I I r 9 i issomonnores e 15.3-12
1 t za , J o -
. I
\
1 i 15.3.4 Leeked Reter'with L.s s of Offsite Power ' 4 The looked-roter event concurrent with a less of offsite power at Mme of generator trip was.also evaluated. A locked-roter event with reactor { I trip-turbine trip &nd subsequent generator trip would ultimately result in 4
,1 coastdown in the reacter coolant flow due to tripping of the reacter coolant ptgs.
Newever, by the time this effect is seen, the severity of the l t transient will'have turned around. Therefore, the peak clad temperatureg RCS j pressure, and percentage of fuel damage, for the locked ,reter/ shaft break ! event discussed above, are 41 e applicable to the case with a concurrent less i of offsite power and the results need not be presented again,
. , i 15.3.5 Raf3tyggg -
- 1. .
'Byren/Braidweed Stations Updated Final 5sfety Analysis Report (UFSAR),
Decket Numbers 50-454, 455, 458, and 457, December 1988. l
- 2. ;
Burnett, T. W. T., et al., 'LOFTRAN Code Description', WCAP-7907, June i
' 197t. .
i
- 3. Margrove H. C.,
/
- 'FACTRAN - A Fortran - IV Code for Thermal Transisnts ij a 00g Fuel Red *, M -7 M , June 1972.
i
- 4. Shefcheck, 1. S., .
{
" Application of the THINC Program to PWR Cosign," i WCAP-7359-L (Proprietary) August 1969 and WCAP-7838, January 1972.
5.' i Baldwin, W. S., Morrian, M. M., Schenkel, H. S. and Van De Wallta 0. J..
'An Evaluation of Less of Flow Accidents Caused by Power Systen Frequency {
Transients in Westing %euse PWRs,' WCAP-8424. Revision 1 h ne 1975. tesor.ierostaas ' 15.3-13 ,
T d::, , i TABLE 15.3-1 t TIM SEQUENCE OF EVENTS FOR INCIDENTS
- WM!CM R8tXT IN A DECR8A3! IN REACTOR CC-1 ANT SYSTEN FLOW 4
gl@g,7 g
- T!E (sec.)
#artial Less of Forced Reactor Coolant Flow Coastdown begins 0.0 Low flow reactor trip 1.72 l
Rodsbehintodrop 2.72 l, Minimum DNOR occurs 4.10
-Complete Lass of Forced Reactor Coolant Flow l
Four Loop ' Coeration All operating 0.0 pumps lose power and begin coast-ing down Reactor cool- 0.0 ant pump cador- ' voltage trip point reached Rods begin to drop 1.5 Minimum DNBR occurs 3.6 l l l l iseocuvessass 15.3-14
7 3
~4;
.e )
P y' 1 n TABLE 15.3-1(Cont'd)
+
.'~
Affd9gl Eg[ TIME (see.)
- Four Loop t.
Coeration h Reactor Coolant Pwe ! Shaft Break ! J l Puus shaft breaks 0.0 l
! ,i Low flow trip point reached i 0.05 i Reds begin to .
drop i ( +
. 1.05 t Maatsum RCS '
pressure occurs l i 3.25 i MaaiWWE clad 1 temperatura I L I t occurs 3.50 u . i i 1 t i (; i 1- ! l'
, t 1 i 1 I 1; '.
\ e . i 1 r r l I. l 1' - I 1 L : i l r I i b l l h l \, ! tene m essaas 15.3 15 ' r
- i e
TABLE 15.F2 . SIAR4ARY OF RfMLTS FOR 1"MT BREAK in_m!ENT I 3 FOUR LOCPS I i OPERATING 1 s INIT! ALLY Maximus React.or Coolart System . i Pressure (psia) l 2581 i i 1 Maxima Cladding Temperature ('F) I L i Core Het Spot i 1853 i
=Zr-H2 O reaction at core hot spot '
1 (5byweight) ; 0.315 ' l ! d' l l r i f t s t l .e ; l. i i P i i m iwossaas e 15.3-16 ' 1 .. .. . .-. -...:-.. - - - . - - _ . - - . . . ~ _ ~ . _ . - - - . - - . . _ . - . _ _ ._ - _ - _ - - - .. .
J g l; - ~.
.i , !
i 1.4
. 1.3 4
. +8 i= 1 L
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u 1.6 -
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t r I 1.2 ' t
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i b i i
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i
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l-i
- 9. ~ !
e 1 2 3 4 ! 5 4 7 8 9 10 71 4 (SED
- i 1 BYRolyORAIDWOOD STATIONS I 1
FIGURE 15.3 1 FLOW TRAll51DITS F0ft FOUR LOOPS IN OPERATI0ll, TWO PUMPS COAST!ses 00WN k
f . .! a
* . j m ,, $.0 ..
1.2 - 1. t ' j
.8- '
i i
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- l. 2400. -
u i asa. . y , ln... i. g ! 3 2100. f E - l.. l! t 1990. I l , 11 - I 8' 1. 2. 3. 4 S. 6. F. 3. 9. .ie, fles tagg) e i STRINVERAIDWOOD STATIONS t t FIGURE 15.3 2 ! I' NUCLEAR POWER Al@ PRES $URIZER PRESSURE - I- TRAftS!ENTS FOR FOUR LOOPS IN QPERATION, 11f0 PL8FS C04STINB DOWN : 1 t: { t
[ ,
]
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l 1.4 ;
, 1.3 >
- 1. AVERAst CHAISIEL
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i , l-i s. 0 1 2 3 4 5 6 7 a e to { Ytar case .
.. t i, -. ;
BYR0lVBRA!0 WOOD STATIONS FIGURE 15.3 3 l ' AVERAGE Ale NOT CHANNEL HEAT FLUX TRAlts!DITS FOR FOUR LOOPS IN OPEUTI FOUR M N S C045YING 00WN
. . . - - - . . . ~ . - _ ~ - - . - - - _ . - . . . - . _ . - -
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* = > 4 Ting i . 7 . , ,, !
(sets ! i t, t a 1 i i e E i i BYRolVORAIDWD00 STATIONS l FIGURE 15.3-4 DNBR VERSUS TIME FOR FOUR LOOPS IN i OPERATI0ll. TWO PLMPS COASTING DOWN
.. ,- . ,,-.. . . ,.- - . . . ~ . . . . . _ . . - . . - - - . - . . . - - . - - - -
3 g l i
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- e. $
L . i a > . . . , . , t- i. i tras esses ! I N e I e i e f I l-i f t WHOIVBRAIDWOOD STATIONS l: FIGURE 15.3 5 1 i CORE FLOW C0437DOWN FOR FOUR LOOPS IN OPERATILM, FOUR MMPS C0ASTING 00WN '
?
- - _ _ _ . ~ . _ _ _ _ _ _ . _ , _ _ _ _ _ _ _ _ _ . . . . - . _ _ _ - - . - _ - - _ - -
1 -teo j r I 1.4 * .
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I 4. 0 1 2 3 4- ! 5 4 F 4 9 te ' Ting ' (set) ! tite. I s i- 3490. i e 2300. ! I j C ! i 2200. - I l
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i .... . l
's 1969. ,
i 1890.
- 9. 1. 2. 3. 4. 5. 4.
F. 4. 9. 10. TIM (Sitt I BYhtuyBRAIDWOOD STAT!0Its ' F18URE 15.3 4 NUCLEAR M A W PRESSURIZER PRE TRAllSIDf73 FOR FOUR LOOPS Ill OPERAT FOUR PWPS C0ASTIIIB 00WII l
.. . _ . - _ - - - - ..- -_ - . --.~ . - . - . _ ... - -. - - ... - . . -
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l ! o .I ' ' i i b i l S. - 8 1' 2 3 4 5 4 7 8 9 to '
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I. , BYMBVMAllMXW STATIONS I FItuRE 15.3 7 f l AVERAGE Ale WT CNAlelEL NEAT FLUX l TRAll81Difs FOR FOUR LOOPS IN OPERAT!0fi, FOUR MSIPS C0AST!ll8 00181 : I _ ~ . . . _ _ . _ . _ . _ _ _ _ . _ _ _ _ . _ _ _ _ _ . . _ _ _ _ . _ _ _ . . _ - . . . _ . . _ _ . _ . . . . . . . _ _ . . .
- v. p .; , , _ . . . . , . - .
e
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N *! W STATIONS em a. .m* f GWRE 15.3-8
' ONBR VE.Jcs TIME FOR FOUR LOOPS IN I, OPEkATION, FOUR PLMPS C0ASTING 00WN f
r 4 m , 4-
- qw s ' in .,.n..e $ . -
,w,- , -- * * + - - - ~ * - + *~~~~ * ~ ' = ~ ~ ' " = -
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46 . . . -: . . . . e a 4 e a to
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- TAT! Git m
FIguilt 15.3-9 4
'P-FLolf TitAllS!ENT5 FOR FOUR LOOPS IN OPERATI0ll, itEACTOR C00LAlti MMP SHAFT OREAK tg. ,, ,, . . . , . . . - . . - * * * * ' ' ' - ' ' ' ' ' " * ' ' ~ * ' _ ,,.,.,m.. ~ -*W"' '
- .- . - . . - . . . . - - . . - - . . . - - - - - - ~ - - ---- --- - ,
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- I 1
i2600. 2500. / k 1 5 k 2480. f t l 2340. < l g 2ne. .
% ~._ '
I 2,.. t 2000.' O. 1. 2. 3. 4. 5. 4. F. 8. 9. 10. Tins (set) . S M Bl/3RAllR1000 STATIONS FIGURE 15.3-10 WCLEAR POWER AS REACTOR C00LANT PRESSURE FOR JOUR LOOPS IN OPERATION, REACTOR COOLANT P W P SHAFT BREAK
. . _ - . - . . . . - . . - . _ . . = . - . _ . . - . . . . - . - - - - - . . . . - .-
. . . . .. .. ~ . . . . _ _ . . _ - . . .- -- . _ . . - - . _ . . . .
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O. ' I 8 1 2 3 4 5 6 7 s , to ' l:- Ting (sec) i STMBVMAIDWOD STAT!0Its ' 1 l FIGURE 15.3-11 1 l-AVERAGE Ale W T O W9fEL HEAT FLUX TRAIISIDCS FOR FOUR LOOPS Ill 0F.1AT1018, REACTOR C00LAlff M51P SNAFT Batt/A l tt .
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. 800.
l l! 600. ?
-I* I. 3. 4 3. g, y, g, 7tNE tasc) t W
BYRDWBRAIDWOOD STATIONS l FIGURE 15.3 12 ' MAXIfRM CLAD TEMPERATURE AT NOT SPOT FOR FOUR LOOPS IN OPERATION, REACTOR C00UNT MMP SHAFT OREAK i y , ,sw. ,w e- e--,,- ,--,.n,----- ,,--.--ee -, n e .e m - m m
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9 , _ 8 , , 15.4'SPECTRUNOFR00CLUSTERChNTR0LASSEMBLYEJECTIONACCIDENTS 1 15.4.1 Identification of Causes and Accident Description l i 1 i... This accident is defined as the mechanical failure of a control rod mech { pressure housing resulting in the ejection of a rod cluster control assembly $ ('ACCA) and drive shaft. The consequence of this me:hanical failure is a rapid I positive reactivity insertion together with an adverse core power s distribution, possibly loading to localized fuel rod damage. ' 15.4.1.1 Desion Precautions and Protection Certain features are intended to preclude the possibility of rod ejection act.ident, er to limit the consequences if the accident were to occur. These include a sound, conservative mechanical design of the rod housings, together '
-with a thorough quality control (testing). program during assenhly, and a ,
nuclear design which lessens the potential ejection worth of RCCAs, and minimizes the number of assemblies inserted at high power levels.
. / MechanicalDesign, s
The mechanical design is discussed in Section 4.6 of the Byron /3raidwood Stations UFSAR (Reference 1). Mechanical design and quality control procedures intended to preclude the possibility of a RCCA drive mechanism housing failure are listed below: a. Each full length control rod drive mechanism housing is completely assembled and shop tested at 4100 psi.
- b. .The mechanism housings are individually hydrotested after they are
' attached to the head adapters in the reactor vessel head, and checked during the hydrostatic test of the completed reactor coolant system.
c.
' Stress levels in the mechanism are not affectou e
- Mee w
- system transients at power, or by the thermal w omen , a m' " . loops.
Moments induced by the design-basis es-thquake .an be occapted within tasocto/nesass 15.4-1 '
qAt q , ,
]
s f h
, , i
' the allcwable primary working stress range specified by the ASMB Code. I
; Section !!!, for Class.1 components.
d. The latch mechanism housing and rod travel housing are each a single '
. length of forged Type 304 stainless steel. This material exhibits excellent notch toughness at all temperatures which will'he }
encountered. l A significant margin of strength in the elastic range' together with the large energy absorption capability in the plastic range gives additional assurance ', that' gross failure of the housing will not occur. The joints between the latch anchanism housing and head adapter, and between the lat:h mechanism p I housing and rod travel housing, are threaded joints reinforced by canopy type - welds. Adainistrative regulations require periodic inspections of these (and other) welds. l l Nuclear Desion '
<Even-if a rupture of a RCCA drive mechanism housing'is postulated t the -
operation of a pla'nt utilizing chemical shim is such that the severity of an ~ ejected RCCA is'inhersntly limited. 'In general, 'the reactor is operated with
.the RCCAs inserted only far enough to permit load follow. Reactivity chang'es caused by core depletion and xenon transients are compensated by boron .
changes. Further, the location and grouping of control RCCA' banks are selected during the nuclear design to' lessen the severity of a RCCA ejection
- accident. ,
' Therefore, should a RCCA be ejected from its normal position during full power operation, only a minor reactivity excursion, at worst, could be' '
expected to occur. H9 wever, it may be occasionally desirable to operate with larger than normal insertions. For this reason, a rod insertion limit is defined as s' function of power level. Operation with tha RCCA's above this limit guarar'oes adequate shutdown capability and acceptable power distribution. The position of all RCCAs is continuously indicated in the control room. An alarm will occur if a bank of RCCAs approaches its insertion limit or if one RCCA
' issov:to/ososse y 15.4-2
.. _ . - ~ . _ _ . _ _ _ . _ _
- ,. . 1 h
'j0 d
i
'eviates from its bank. Operating instruction require boration at low levei alarm and emergency boration at the low-low alare. .
Reactor Protection I The reactor prote: tion in the event of a rod ejection accident has been
" described in Reference 2. The protection for this accident is provided by .
T high neutron flus trip (high and low setting) and high positive rate neutron flua trip. Effects on Actiacent Housinos , I [ Disregarding the remote possibility of the occurrence of a RCCA mechanism 1 5 L housing failure, investigations have shown that failure of a housing due to either icsgitudinal or circumferential cracking would not cause damage to ,. i o' adjacent housings. However, even if damage is postulated, it would not be espected to lead to a more severe transient since RCCA's are inserted in'the t I core in symmetric patterns, and control rods immediately adjacent to worst ; 4
. ejected rods are not in the core when the reactor is critical. Damage to an 3- {
adjacent housing could, at worst, cause that RCCA not to fall on receiving a' { trip signal; however, this is already taken into account in the analysis by i
. assuming a stack rod adjacent to the ejected rod.
- 1 E i 15.4.1.2' Limitino Criteria '
l This. event it classified as an ANS Condition IV incident. See Subsection 15.0.1 of the Byron /Braidwood Stations UFSAR for a discussion of ANS classifications. Due to the entremely low probability of a RCCA ejection accident, some fuel damage could be considered an acceptable consequence. i L. - Comprehensive studies of the threshold of fuc1 failure and of the threshold or significant conversion of the fuel thorac 1 energy to mechanical energy, have l been carried est as part of the SP8RT projact by the Idaho Nuclear Corporation l (Reference 3). Extensive tests of U02 zirconium clad fuel rods representative l of those in pressurized water reactor type cores have demonstrated failure { l l 1 l u issov:1o/osuss 15.4-3
= .e., m . . , ....~..m.--.-_,,,,,,.-,,,..,.-.-..~._.,~.r- , , , . , - . , - , . -m..,,_.r . . _ , - v w-- -
-' ^
n ..
. m ,
F bi, thresholds in the range of 240 to 257 cal /ge. However, other rods of a slightly different' design have ethibited failures as low as 225 cal /ge. These results, differ significantly from the TREAT (Reference 4) results, which c indicated a failure threshold of 280' cal /ge. Limited results have indicated H that this threshold decreases by about 105 with fuel burnup. The clad failure mechanism appears' to be moltin; for zero burnup rods and brittVe fracture for ) L
. irradiated rods. Also important is the conversion ratio of thermal to h
i mechanical energy. This ratio becomes marginally detectable above 300 cal /gm n
" for unirr6diated rods and 200 cal /gm for irradiated rods; catastrophic ;
failuro, (large fuel dispersal,plarg' spressure rise) ' event for irradiated rods,'did not occur below 300 cal /ge. : In view of.the above experimental results, criteria are applied to ensure that thoro is little or no possibility of fuel dispersal in the coolant, gross , lattice distortien,,or severe shock waves. - These criteria are: . a.
~ Aversgo fuel pellet enthalpy at the hot' spot below 225 ca)/gm for unirradiated fuel and 200 cal /gm for' irradiated fuel.
l
- b. Average clad temperature at the hot spot below the temperature at I which clad embrittlement may be expected (2700*F).
l. 7 c. Peak reactor coolant pressure less than that which could cause stresses to exceed the faulted condition stress limits.
- d. Fuel molting will be limited to less than ten percent of the fuel volume at the hot spot even if the average fuel pellet enthalpy is ,
below the limits of criterion & above. ' 15.4.2 Analysis of Effects and Consecuences Method of Analysis f The calculation of the RCCA ejection transient is performed in two stages, first an average core channel calculation and then a hot region calculation. 1ssov.1onsaias 15.4-4 1
..m
F : p . , o t H y - i i L I The average ' core calculation is performed using spatial neutron kinetics !- i
'. methods to detamine the average power generation with time including the ;
' 'various total core fe'edback effects, i.e., Doppler reactivity and moderator { reactivity. Enthalpy and temperature transients in the hot spot ere then 1, determined by multiplying the average core one gy generation by the hot ; b channel factor:and performing a. fuel rod transient heat transfer calculation. . The power.distributhn calculated without feedback is possimistically assumed R .to persist throughout the transient. '!' h,' A detailed discussion af the method of analysis can be found in R6forence 2. Aversoe Core Analysis }' ' , , t.g . The spatial kinetics computer code, TWINKLC (Reference 5)', is used for the. ' average core transient analysis. This code solves the two group neutron .; diffusion theory kinetic equation in one two or three spatial dimensions ' L
'(rectangular. coordinates) for six delayed neutron groups and up to 2000 spatial' points. The computer code includes a detailed multiregion, transient -
fuel-clad-coolant heat transfer model for calculation of pointwise Doppler and moderator feedback effects. In this analysis, the code is used as a one
. dimensional axial kinetics' code since it allows a more realistic ,
representation of tho' spatial effects of axial moderator fesaback and RCCA * } movement. However, since the radial dimension.is missing, it is still necessary to employ very. conservative methods (described in tho'following) of eticulating the ejected rod worth and hot channel factor. ' e i< Hot Soot Analysis In the hot spot analysis, the initial heat flux 1.6 equal to the nominal times [ the design hot channel factor. During the transient, the heat flux hot , cha'nnel factor is linearly increased to the transient value in 0.3 second, the time for full ejection of the rod. Therefore, the assumption is unie that the
- hot spots before and after ejection are coincident. This is very conservative since the pean after ejection will occur in or adjacent to the assembly with
?
1ssovao/ososes . 15.4-5
% 5 V
i
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. . ..;. . . . + . -
f j,
,)*
t the ejected rod, and prior to ejection the power in this region will i c necessarily be depressed.
'[ The hot spot analysis is performed using the detailed fuel-and cladding '
transient heat transfer computer code, FACTRAN (Reference 6). This computer code calculates the transient temperature distribution in a cross section of a metal clad 002 : fuel rod, and tho' heat flux at the surface of the rod, using s j as input the nuclear power versus time and the local coolant conditicas. The zirconium-water reaction is' explicitly represented, and 911 material , properties are represented as functions of temperature. A conservative pellet : radial power. distribution is used within the fuel rod. L 'FACTRAN uses the Dittus-Boelter or Jens-Lottes correlation to determ
- film heat transfer before DNB, and thw Bishop-Sandberg-Tong correlation (Reference 7) to determine the film boiling coefficient after DNB. The 85T I
. correlation'is conservatively used assuming zero bulk fluid quality. The DN8 L .
L' ratio is not' calculated, instead the code is forced into DN8 by specifying a ' L conservative DNS heat flux. The gap heat trancfor coefficient can be - calculated by,the crie; however, it'is adjusted in order to force the full power steady-stath temperature distribution to agree with the fuel heat ' L transfer' design codes. L . y , Systen C.. .ssure Analysis , L 1 Because safety limits for M 1 damage specified earlier are not exceeoed. there is little likelihood of fuel dispersal inte the coolant. The pressure i
. surge may therefore be calculated on the basis of conventional heat transfer '
from the fuel and prompt heat generation in the coolant. The pressure surge is calculated by first performing the fuel heat transfer calculation to determine the average and hot spot her. flux versus time. Using this heat flux data, a THINC (Reference 8) calculation is conducted to determine the volume surge. Finally, the volume surge is slaulated in a plan L transient computer code. l
, This code calculates the pressure transiert taking tasov:1o/osoJst 15.4-0 l . S,._.# .. .,. -
~
b into account fluid transport in the reactor coolant system and heat transfer to the s. team generators. No credit is taken for the possible pressure reduction caused by the assumed failure of the control rod pressure housing. 15.4.2.1 Celeulation of Basic Parameters Input parameters for the analysis are conservative y1' selected on the basis of values calculeted for this type of core. The more important parameters are discussed below. Table 15.4-2 presents the parameters used in this analysis. Docted Rod Worths and Hot Channel Factors The values for ejectsd rod worths and hot channel factors are calculated using either three dimensional static methods or by a synthesis method employing one dimensional and two dimensicnal calculations. Standard nuclear % sign codes dre used in the analysis. No credit is taken for the flux flattening effects of reactivity feedback. The calculation is performed for the maximum allowed bank insertion at a given power level, as determined by tne rod insertion limits. Adverse xenon distributions are considered in the calculation. Appropriate margins are added to the ejected rod worth and hot channel factors to account for any calculational uncertainties, including an allowance for nuclear power peaking due to densificatien. Power distributiot.s before and a*ter ejection for a " worst case can be found in Reference 2. During plant startup physics testing, ejected rod worths and power distributions are measured in the zero and full power rodded configurations and compared to values used in the analysis. It has been found that the ejected rod worth and power peaking factors are consistently overpredicted in the analysis. Reactivity Feedback Weichtina Factors The largest temperature rises, and hence the largest reactivity feedbacks occur in channels where the power 13 higher than average. Since the weight of 1ssov:icuossa 15.4-7 ' K
-. . ~ . . - - - - - . - . - - - . . . - -- . - - ..
,s j
O .. .
'a region is' dependent on flux, these regions have high weights. This means
'l
.that the reactivity feedback is larger than that indicated by a simple channel l
analysis. Physics calculaticas have been carried out for temperature changes !
'withLa flat tamperature distribut, ion, and with a large number of axial and radial temperature distributions. Reactivity changes were compared and effective weighting factors determined. These weighting factors take the form i
of multipliers which when applied to single channel feedbacks corr 9ct them to } effective whole core feedbacks for.the appropriate flux shape. In this i analysis, since a one dimensional (axial) spatial kinetics method is employed.
' axial weighting is not necessary if the initial condition is made to match the '
ejectedrodconfiguration. In addition, no weighting.is applied to the l moderator feedback. .A conservative radial weighting factor is applied to the l 1
-transient fuel tesperature to obtain an et fective fu61 temperature as a ;
function'of time accounting for the missing spatial dimension. These
. weighting fa: tors have also been'shown to be conservativs cumpared to three :
' dimensional: analysis (Referer.co2).
9 Moderator and Donoler Coefficitt.t ' The critical boron concentrat' ions at the beginning of life and end of life are L [ adjusted in the nuclear code in order to obt:ain moderator density coefficient curves which are conservative compared to actual design conditions for the plant. As discussed above, no weighting factor is applied to these results. The Doppler reactivity defect is determined as a function of power level using - a one dimensional steady-state computer cods with a Doppler weighting factor i ,. of 1.0. The Doppler defect used is 0.91% ao for beginning of life cases and L 0.64%'ao for the end of life cases. The Doppler weighting factor will Y, increase.under accident co'nditions, as discussed abovo. Delayed Neutron Fraction.J . Calculations of the effective delayed neutron fraction (8,ff) typically yield values no less then 0.70% at beginning of life and 0.50% at end of 1 He C for the first cycle. The accident is rensitive to 6 if the ejected rod
- worth is equal to or greater than 6 as in zero power transients. In order tesovao/oscass 15.4-8
-- x
- g. ,
j lc '
. 1 m j to allow for future cycles, pessimistic estimates of a of 0.555 at begir.ning 1
of. cycle and 0.445 at'ond of ' cycle were used in the analysis,
]
j )- Trio Reactivity Insertion ' L ; i The trip reactivity insertion assumed 1 .Jven in Table 15.4-2 and includes I the effect of one stuik Per.A. These VGw are redut:ed by the ejected rod reactivity. The shutdown reactivity was simulated by dropping a rod of the l requiied worth into the core. The start of rod moti:n occurred 0.5 seconds after the high neutron flux trip point was reached. This 61ay is assumed to consist of 0.2 seconds for the instrument channel to produce a signal 0.15 seconds for the trip breaker to open and 0.15 seconds for the coil to release p the rods. A curve of trip rod insertion versus time was used which assumed l
+
that' insertion to the dashpot does not occur until 2.7 seconds after the stut y of fall. The choice of such a conservative insertion rate means that there is-
'over enn second after the trip point is reached before significant shutdown reactivity is inserted into the core. This is a particularly important contervatism for hot full power accidents.
The minimuc design shutdom available for this plant at HZP ray be. reached l( only at end of life in the equilibrium cycle. This value includes an [ allowance for the worst stuck rod, adverse menon distribution conservative
' Doppler and moderator defects, and an allowance for calculational
~
uncertainties.. Physics calculations for this plan haveshown'that the effect of two stuck RCCAs (one of which is the worst ejectad' rod) is to reduce the ! r
' shutdown by about an additional one percent Ak. Therefore, following a reactor trip resulting from an RCCA ejection accident, the reactor will be L
suberitical when the core returns to HZP.
~
Reactor Protection Reactor protection for a t ad ejection 1. provided by hiCh neutron flux trip ' (high and bw setting) and high pcsitive rate neutron flux trip. Tiese protection functions 6re part rf the reactor trip system. No singla failure iss:vno/ososes 15.4-9
. , < , - - e . . . , _ . . , . . . , . . _ , , - . . . . . . - . . . . ., . _ _ . - . , , , , , - . , , _ . . . , . , , , -
g 14 .. . u- , 4 l t i
)
of the reactor' trip system will negate the protection functions required for , the rod ajection accident, or adversely affect'the consequences of the-accident. , W. 15.4.2.2 Results ,
~Csses are presented for both beginning and and of life at zero and full power.
- 1. Seeinnine of Cycle.' Full P. m 7
Control bank D was assumed to be inserted to its insertion limit. The q worst ejected ~ rod worth and hot channel factor were' conservatively I calculated to be 0.255 Ak.and 6.10, respectively. The peak hot spot i L clad average temperature.'was 2351'F. The peak hot spot fuel center j temperature. reached molting,; conservatively, assumed at 4900*f. However, molting.was restricted to less' tilan 105 of the pellet.
'2. Becinning of Cycls. Zero Power- I l
For;this condition, control bank 0 was assumed to be fully inserted and g " banks B and C were at their insertion ' limits. The worst' ejected rod is I located in control bank D and has a worth of .7655 Ak ano a hot channel factor of 11.5. The peak' hot spot clad average temperature reached 2274'F, the fuel center temperature was 3585'F. '
- p. 3. End of Cycle. Full Pgg l
l Control. hank D was assumed t'o be inserted to its insertion limit.The ' ejnted rod worth and hot channel factors were :enservativsly calculated
~
to be 0.25% ok and 6.40, respectively. This resulted in a clad average ' temperatur3 of 2187'F. The peak hot spot fuel temperature reached meltbg conservatively assumed at 4800*F. However, molting was restricted to less than 105 of the pellet. l l 1 l
, issov:to/osesse 15.4-10 u - .-- - - _ . _ --- - ----- - -.- - -- -- - -'- ~ ~~~ ~~~ ~~~- ~ ~
[. , , Yl- *
'j E- 4. End of Cx le. Zero Power y
7 -) The~ ejected rod worth and hot channel factor for this case were obtained ' assuming control bank D,to be fully ins 9rted and banks C'and 8 at its 1nsertion limit. The results were .905 Ak and 21.0, respectively. The peak clad average and fuel center. temperatures- were 2603 and 3893'F.The ' Doppler' weighting factor for this case is si p ificantly higher than for 4
,n the other cases due to the very large' transient hot channel factor.
A summary of the cases presented above is given in Table 15.4-2. The nuclear ; power and hof spot fuel and clad temperature transients for the worst cases. [ ,
.are presentediin Figures 15.4-1 thrt, ugh 15.4-4. (Beginning of life full power l and and of life zero power.) 1 1
.i The calculated sequerce of events for the worst case rod. ejection accidents, ;
as'shown in Figures 15.4-1'through 15.4-4, is presented.in Table 15.4-1. For all cases, reactor trip occurs very early in the transient after which the ! nuclear power excursion is terminated. The reactor will remain suberitical ' following reactor trip. p , y The ejection of an RCCA constitutes a break in the reactor coolant system, located.in.the reactor pressure vessel head. 'The effecte and consequences of ' js ' loss of coolant accidents are discussed elsewhere in this report. Following the RCCA ejection, the operator would follow the same emergedcy instructions a
'as for any other ' ass of coolant accident to recover from the event ,
t
- l. Fission Pr~1uct Release It is' assumed that fission products are released from the gaps of all rods entering DNB.
In all cases considered less than 105 of the rods entered DNB based on a detailed three dimensional THINC analysis (Reference 2). (
\~
i naovao/ososse [, 15.4-11
v' x ..: - (( 4 l x , i
) O , a .
Pressure Surne i , A detailed niculation of the pressure surge for an ejection worth of one dollar at beginning.of life, hot full power, indicates that the peak pressure
< does not exceed that which would cause stress to exceed the faulted ;
stress' limits (Reference 2). Since the severity of the prose.it analysis does not exceed'the " worst ca4e* analysis, the accident for this plant will not - t result in an excessive pressure rise or further damage to the reactor coolant . ; system. ' ' i Qtii,ce Defc,rmations L i L A large temperature gradia.it will exist in the region of the hot spot. Since ; the fuel rods are free to move in the vertical direction, differential - expansion between' separate rods cannot produce distortion. However, the i temperature gradients across individua.1 rods may produce a differential c expansion tending te bow the midpoint of the rods toward the hetter side of l , the rod. L Calculations have indicated that this bowing would result in a negative reactivity effect at the hot spot since Westinghouse cores are under-moderated, and bowing will tend to' increase the under-moderation at the hot spot. Since the 17 x 17 fuel design is also under-moderated, the same effect would be observed. In practice, no r,igMficant bowing is anticipated, L' since the structural rigidity.of the core is more than. sufficient to withstand
~the forces produced. Boiling in the hot spot region would produce a not flow away from that region. However, the heat'from the fuel is released to the i
water relatively slowly, and it is considered in:onceivable that cross flow will be sufficient to produce significant lattice forces. Even if massive and rapid boiling, sufficient to distort the lattice, is hypothetically postulated, the large void fraction in the hot spot region would produce a reduction in the total core moderator to fuel ratio, and a large reduction in this ratio at the hot spot. The not effect would therefore be a negative feedback. It can be concluded that no conceivable mechanism exists for a n positive feedback resulting from lattice deformation. In fact, a small negative feedback may result. The effect is conservatirely ignored in the anclysis. issov:10/osessa 15.4-12 I e ,; . a = - - - - -- - - - - - -
f Ti . 15.4.3 Conclusions, ' L
. Conservative analyses indicate that the described fuel and cladning limits are h
I not exceeded.. It is concluded that there is no danger of sudden fuel dispersal into the coolant. Since the peak pressure does nat exceed that L j o' , .
' which wouid cause stresses to exceed the faulted condition stress limits, it I is concluded that there is no danger of further consequential damage to the reactor coolant system. The analyses have demonstrated that the fission j
1 product release, as a result of a number of fuel rods entering DNS, is limited
'to less than 10% of the fuel rods in the core.
y 15.4.4 References r i
'1.
L -" Byron /Braidwood Stat' ions Updated Final Safety Analysis Report (UFSAR) " Docket Number 50-454, 455, 456, and 457, December 1988. ' 2. Risher, D. H., Jr., "An Evaluation of the Rod Ejectirn Accident in
)
Westinghouse Pressurized Water Reactors U:ing Spatial Kinetics Methods," ' WCAP-7588, Revis' ion 1-A, January 1975. 3. Taxelius, T. G. (F.d), " Annual Report - SPERT Project October 1968, l j September 1969," Idaho Nuclear Corporation IN-!370, June 1970. i 4
- 4. Liimataninen, R. C. and Testa, F. J., " Studies in TREAT of i Zircaloy-2-Clad, UO 2 -Core Simulated Fue.1 Elements" ANL-7225, January -
June 1965, p. 177. November 1966.
- 5. Risher, D. H., Jr. and Barry, R. F., " TWINKLE - A Nulti-Dimensional Pautron Monetics Computer rode," WCAP-7979-A (Proprietary) and WCAP-8028-A l (Non-Proprietary), January 1975.
- 6. Hargrove, H. G.,
"FACTRAN - A Fortran-IV Code for Thermal Transients in a 002Fuel Rod,* WCAP-7908, June 1972.
1sm:io/osanas 15.4-13
- , +__
n, g ,
- 2.,. . ..
, i
' I
!. 1 B
i , 9
- a. .
- 7. Sishop, A. A., Sanbrg, R. O. and Tong, i., S., " Forced Convection Heat
. Transfer at High Pressure After the Critical Heat Flux," ASME 65-HT-31,
- August.1965. ;
- 4. Shefcheck, J. S.. i
" Application of the THINC Program to PWR Design," i
. . WCAP-7359-L (Proprietary) August 1969 aad WCAP-7838, January 1972. .
n-i 4 t [.
. .c' .
p. p P
- e l
k r
't I
Issov:io/ososas 15.4-14
-w'-- ~- w - e --+,.,-,~.---..--s
, , . - - - , - - . - - - ,n- -, . Aa , ,
M , j f' if , , 3
- TABLE 15.4 1 I l
TIME SEQUENCE OF EVENTS FOR THE SPECTRUN ! QF RCCA EJECTION EVENTS n , ACCIDENT TIME EfjN1 Isoc.) < Red Cluster Control ' Assembly Ejection . . 1.. Beginning-of- . Initiation of rod ejection
! Life, Full Power 0.0 i Power range high neutron flux 0.05 , setpoint reached Peak nuclear power occurs 0.14' Rods begin to fall into core 0.55 Peak fuel average temperature 2.28 occurs ,,
Peak clad temperature occurs 2.35 Peak heat flux occurs 2.36 [ 2. End-of Life Initiation of rod ejection Zero Puwer . 0.0 , Power range high neutron flux l low setpo'nt reached O.18 t Peak nuclear power occurs l 0.21 , Rods begin to fall into core 0.68 L Peak clad temperature occurs 1.55 Peak heat flux occurs 1.55 Peak fuel average.tempferature 1.80 ' occurs iesov:1c/osossa
- j. .
15.4-15 R + - -v+g @ fd tWgr eer'hr'"'TWW"'"WWVT','PT*P' N" ' * *ha 8'*D""WN '"-''" F'-"'"' 8'8'***N'"-'*F" "-N'"'^ '^ --""--"--- - - --
< ~ ~.
4 TABLE 15.4 * . 1, PARAIETERS USED IN TK AllALYSIS OF TK A00 CLil5TER CD11 TROL
- ASSEISLY EJECTIGI ACCISENT 6
" SOL-NP BIR.-IIZP EE.-NP TIE IN LIFE BEGilOIIIE SEGIISIIIE ECL-HZP i-EIS ' EW ' i Power level, % 102 i 0 102 3 ' ! Ejected red worth, X K 0.25 i 0.765 0.25 0.90 ' Delayed neutron fraction, X - 0.55- 6.55 ' 9.44 0.44 n Feedeck reactivity weighting 1.30 2.07 i 1.30 3.55 ; Trip reactivity, % K 4.0 2.0 4.0 2.0 ! F,before rod ejection 2.60 -- 2.60 -- i F,after rod ejection 6.10 11.5 6.40 t 21.0 L Ilumber of operational pumps 4 2 ! 4 2 ' Itax. fuel pellet average 4136 3075 ! temperature, 'F 3843 3424 ~ , t Ilan. fuel conter l >4900 3585 >4000 temperature. *F 3893 - l Max. clad average 2351 temperature. *F 2274 - 2187 PSS3 . 4 Max. fuel store energy, cal /gm 182 128 ' 1 166 145 -
% Fuel Melt <101 '
0 <105 0' i
- iseernwesens 15.4-16
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TIMC (SCCl t a b y t i ff a BYR0lVBRAIDWOOD STATICNS ' FIGURE 15.4-1 NUCLEAR POWER TRANS!ENT pSOL HFP RCD EJECTION ACCIDENT l t . '. f,l, .I
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r, l 1 l i ~ ( , l p BYI10N/3RAIDW000 5 TAT 10NS l FIGURE 15,4-4 H07 SPOT FUEL AND CLAD TEMPERATUDEs u VERSUS TIME, E0L NZP R00 EJECTION ACCIDENT g 3$
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l: 1 l- . 1 l, I u l .. L l' % Attachment 4 I
. LOCA ACCIDENT ANALYSIS
/ . ll , FOR THE BYRON /8AAIDN000' STATIONS UNITS 1 AND 2 l TRANSITION TO WESTINGHOUSE 17X17 VANTAGE 5 FUEL'ASSEM8 LIES ~ l l* e ; L q l I I l I 1 4 i L. 1 L , l, L l' l . l l l '. l :. i '5 aants.e aans19 i w 9 4 w ,ee--o-,,ww- w v e- v--e w w v-e--ww--ow-,e-w,-ve, ,e-weer
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6.2.1.5.3 Containment' Volume 1 Thevolumeusedinltheanalysisis3.10x106 3 ft . - 6.2.1.5.4 Active Heat Sinks The containment ' spray system and the containment atmosphere recirculation I system fan coolers operate to remove heat from the containment. g \ A large break LOCA analysis has been performed for the Byron /Braidwood J Stations incorporating revised fan ~ cooler and containment spray initiation times (see Table 6.2-54). The analysis is based upon a fan cooler minimum
- l essential service cooling water temperature of 45'F. The analysis shows that the worst large break LOCA occurs for the case with maximum RCS Temperature i and produces a peak clad temperature of 1883'F, which is below the acceptable 2200*F limit. {
A curve of the temperature versus heat load is providad in Figure 6.2-25 for , the estimated capacity of one fan cooler. I The sump tesperature was not used in the analysis because the maximum peak ; cladding temperature occurs prior to initiation of the recirculation phase for the containment spray system. In addition, heat transfer between the sump water and the containment vapor space was not considered in the analysis. 6.2.1.5.5 Steam-Water-Nixina ' Water spillage rates from the broken loop accumulator are determined as part of the core reflooding calculation and are included in the containment (C0CO) l code calculation model. 6.2.1.5.6 Passive Heat Sinks The passive heat sinks used in the analysis and their thermo physical properties are given in Table 6.2-55. 1s20v:1o/0:24at-1 1
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[, , Table 6.2-55 was generated by calculating conservatively large' areas and thicknesses.. A complete and detailed list of surface areas and thicknesses of i structures and equipment in the containment was compiled. An uncertainty of . from +10 to +255 was assigned to'each . calculated area. - The containment wall area, which was assumed to have 0% uncertainty, was increased by 105 in order to ensure conservatism. The values in Table 6.2-55 provide a conservative. 4 (high) estimate of the containment heat sinks for use in the minimum ~ l containment pressure analysis, a t j j l l l
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1 TA8LE 6.2 - 51 DECLG BLOWOOWN NASS AND ENERGY RELEASES (CD=0.6) !
? *' , Time gas, gjo, -
Energy Flow IS'C) ' (Lb/Sec) (Btu's/Sec)x10-6 0.0 9.521 5.329-0.05 65,995 36.490 0.20. 66,412
# 36.711 . ,
1.0 64,474 35.860 2.0 54,777 30.975 3.0 43,349' 24.886 4.0 s 33,264 20.860 1 5.0 30,888 18.575 6.0 29,148
' 17.726 !
7.0 27,499 16.985 8.0 24.060 15.518 '! 9.0 20,792 13.904 10.0 18,063 { 12.315 12.0 l 14.494 9.849 14.0 11,383 8.121 , 16.0 7,918 6.181 - , 18.0 5,222 4.543 20.0 3,304 2.863~ 22.0 3.862 2.329 24.0 5.869 2.515 ' 26.0 - 5,017 1.727 28.0 5,667 1.421 30.0 -224 -0.2624 , 30.2099 -245 ) 0.2868 isaov:to/ossass 6.2-180
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l TABLE 6.2 - 52 DECLGREFLOOD.MASSANDENEkGYRELEASES(CD=0.6) l
. Time Nass Flow Energy Flow i R (Lb/Sec) (8tu's/Sec) 44.807 0.0
' 0.0 45.432 0.025068
" 32.4 ~ l 45.632 0.02518- 32.55 45.732 0.02523 32.62 45.932 0.02533 32.75 46.232 0.238 -
! 309.71 56.222 3332.29 413,681.49- . 71.472 249.94- '194,620.70 ! 89.972 267.87 192,573.34 I 110.472 278.49 187,726.45. 132.472 -286.83 182,015.46 155.972 294.45 175,853.68 ' i l l r p.
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BROKENLOOPACCUNULATORINJECTIONSPILLTOCONTA!NNENT(C0 - Time Mass Flow Energy Enthalpy
.15,1g1., (Ls' /Sec) (8tu/Sec) (8tu/1b) 0.0 2,751.25 164,029.64 59.62 >
1.00 2.545.83 151,782.45 59.62 2.00 2.382.73 142,058.64 59.62 3.00 2,248.25 134,040.93 59.62 4.00 2,134.98 127,287.95 s 59.62 5.00 2,037.61_ 121,482.21 59.62 6.00 1,952.2). 116,395.88 59.62 7.00 1,876.47 111,874.96
' 59.62 8.00 1,808.49- 107,822.01 59.62 9.00 1,747.17 104,166.36 59.62 10.00. 1,691.61 100,853.93 59.62 11.00. 1,641.03 97,838.38 59.62 i 12.00 1,594.71 95076.44 59.62 13.00 1,552.02 92531.39 59.62 -
14.00 1,512.53 90177.18 59.62 15.00 1,475.81 87987.80 59.62 16.00 1,441.74 85956.70 59.62 17.00 1,410.02 84065.28 59.62 18.00 1,380.35~ 82296.33 59.62 19.00 1,352.62 80642.97 ' 59.62 20.00 1,326.68 79096.73 59.62 22.00 1,279.39 76277.20 59.62 24.00 1,236.61 73725.39 59.62 26.00 1.197.91 71419.27 59.62 28.00 1,163.06 69341.74 59.62 29.00 1,146.83 68374.29 59.62 issov:1o/ossass 6.2-182
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TABLE 6.2-54 ACTIVE HEAT SINK DATA i FOR NININUM POST LOCA CONTAINNENT PRESSURE
.I. Containmenti Spray System Parameters x '
A. Maximum Spray System Flow' Total 8525 gpm B. Festest post-LOCA initiation of spray system assuming offsite power loss at start of LOCA 35 sec II. Containment' Atmosphere Recirculation Fan Coolers t
.. A. Maximum Number of Fan Coolers Operating 4
'B. Fastest post-LOCA initiation ass' uming offsite power loss at start of LOCA 15 see C. Performance Data See Figure 6.2-25 for Fan Cooler Temperature versus heat load curve.
isaov:to/osaass 6.2-183
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8/8-UFSAR s i e , .; TABLE 6.2-55 PASSIVE HEAT $1NK DATA FOR WININUM POST LOCA CONTAINMENT PRESSURE A. Heat Sink Description
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Material Slab Slab Thiekness . Surface Number Descriotion Waterial (ft) Area (ftt) 1 Structural Steel Carbon 3 teel 0.02083 212107.98 2 Structural Steel Carbon Steel 0.25583 320.84 3 Structural Steel Carboe Steel 0.13250 88.23 i j 4 Structural Steel Carbon Steel 0.19792 299.25 ! 5 Structural Steel Carbon Steel 0.20833 892.5 6 Structural Steel Carbon Steel 0.23958 782.25 L : 7 ' Structural Steel Carbon Steel 0.12500 1107.75 8 Structural Steel Carbon Steel 0.10400 '906.09 l- 9 . Structural Steel Carbon Steel 0.04167 42144.90 10 ! Structural Steel Carbon Steel 0.02500 44799.19 l 11 Structural Steel Carbon Steel 0.16667 531.82 12 Structural Steel Carbon Steel 0.18750 131.67 13 Internal Concrete i Concrete 1.0 101391.04 14 Internal Concrete Concrete 1.0 14766.67 1 1 g 15 Containment Floor Concrete 0.5 L L Containment Floor Steel 0.03362 828.13 I Containment Floor Concrete O'. 5 16-19 Foundation and Sump Concrete ' 0.5 Foundation and Sump Steel 0.02292 1850.20 Foundation and Sump Concrete 0.5 1 I isaov:1o/ossass 6.2-184 l l l
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' B/B-UFSAR I
c TABLE 6.2-55(Cont) PASSIVE HEAT $1NK DATA FOR WININUM POST LOCA CONTAINNENT PRESSURE \ ' Material Slab Slab-i Thickross Surface Number- Description Material (ft) Area (ft2) __ Foundation and Sump Concrete 0.5 Foundation and Sump Steel 0.01563 10134.80 Foundation and Sump Concrete 0.5 ! l Foundation and Sump Concrete 0.5 Foundation and Sump Steel ) 0.04899 23489.55 Foundation and Susp Concrete 0.5 Foundation and Sump Concrete 0.5 Foundation and' Sump Steel 0.15276- 3022.63 Foundation and Sump Concrete 0.5 20 Containment Wall Concrete 0.5 Containment Wall Steel 0.028083 115872.75 Containment Wall Concrete 4.0 ' l-L inaov:1onsanas 6.2-184A v .- g-. - - , , , -,,----c,,.-_,.,m.,,m--n..,.,,-. . , , a-
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t B. Thermophysical Properties Thermal l Density Specific Heat l 3 Conductivity Ib/ft Btu /lb 'F Btu /hr-ft 'F j Concrete 145 0.156 0.92
. Carbon Steel 490 0.12 27.0 -l l
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^ BYRON /BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT l l
Figure 6.2-24 ! Containment Pressure l DECLG (Cg=0.6) Minimum Safeguards NominaT Thot (619.3 Deg-F) 1 1 \< l
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UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6-25 Reflood Transient Core Inlet Velocity DECLG (Ca=0.8) Minimum safeguards Reduc 5d Thot (600.0 Deg -)
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B/B-UFSAR 15.6.4 Spectrum of BWR Steam System Pipino Failures outside of Containment ji :This section is'not applicable to the Byron /Braidwood Stations. . g J 15.6.5 Loss of Coolant Accidents Resultino From a Spectrum of Postulated Pioino Breaks Within the Reactor Coolant Pressure Boundary 15.6.5.11 Identification-of Causes and Frecuency Classification I A LOCA is.the result of a pipe rupture of the RCS pressure boundary. For the l analyses reported here, a major pipe break (large break) is defined as a rupture 2with a total-cross sectional area equal to or greater than 1.0 square -!
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foot (ft ), .This event is considered an ANS Condition IV event, a limiting l < fault, in that it is not expected to occur during the lifetime'of the plant but is postulated as 6 conservative design basis (see Subsection 15.0.1). A minor pipe break (small break), as considered in this section, is defined as a rupture of the reactor coolant pressure boundary (see Section 5.2) with a total cross-sectional area less than 1.0 ft2 in which the normally operating 1 charging system flow is not' sufficient to sustain pressurizer level and
. pressure. .This is considered a Condition III event, in that it is an i L
infrequent fault which may occur during the life of the plant. For .small break LOCAs, the most limiting single. active failure is of an emergency power train'which results in loss of one complete train of ECCS components. The minimum delivered ECCS flow available to the RCS is based on this single failure. For large break LOCAs, the most limiting single failure for purposes of LOCA analyses is the one which produces the lowest containment I pressure. The lowest contain.nont pressure would be obtained only if all containment spray pumps and fan coolers operated subsequent to the postulated { L LOCA. Therefore, for the purposes of large break LOCA analyses, the most l L iimiting single failure would'be the loss of one RHR pump. However, the large break LOCA analyses assume both maximum containment safeguards (lowest containment pressure) and minimum ECCS safeguards (the loss of one complete ' train of 20C5 components), which results in the minimum delivered ECCS flow available to the RCS. 1:20v:1D/062449-11
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.TheacceptancecriteriafortheLOCAisdescribedin10CFR50.46(2)as i
follows: , ", -
- a. .The calculated peak fuel element clad temperature is below the.
requirement of 2200'F. l b. The amount of fuel element cladding that reacts chemically with water i E or steam does not exceed 1% of the hypothetical amount that would be ; generated if all of the zirconium metal in the cladding cylinders Li {
+ surrounding the fufl, excluding the cladding surrounding the plenum 1 volume, were to react.
- c. The clad temperature transient is terminated at a time when'the core l geometry is still amenable to cooling. The localized cladding j.
t- - ' oxidation limits of 17% are not exceeded during or after quenchin'g. J d. The core remains amenable to cooling during and after the break. ; l
- e. The ccre temperature is reduced and decay heat is removed for an extended period of time, as required by the long lived radioactivity remaining in the core. !
L These criteria were established to provide significant margin in emergency j-core cooling system (ECCS) performance following a LOCA. Reference 3 presents a recent study in regard to the probability of occurrence of RCS. pipe ruptures. The quantity of ECCS flow delivered to the unbroken loops, as assumed in the limiting break calculations, is given in Figure 15.6-52 for the condition III (small break) LOCA and in Figure 15.6-26 for the condition IV 4 (large break) LOCA. In all cases, small breaks (less than 1.0 ft2) yield results with more margin to the acceptance criteria limits than large breaks. 1:20v:10/0531ss-12
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15.6.5.2' Secuence of Events and Systems Ooorations L, lShould a major brsak occur, depressurization of the RCS results in a pressure decrease <in the pressurizer. The reactor trip signal subsequently occurs when the pressurizer. low pressure trip setpoint is reached. A safety injection
. signal is generated when the appropriate.setpoint.is reached. These counter-measures will limit the consequences of the accident in two ways:
- a. !
Reactor trip and borated water injection complement void formation in l" causing rapid reduction of power to a residual level corresponding to fission product decay heat. However, no credit is taken in the LOCA analysis for boron content of the injection water. In addition, the 'f insertion o'f control rods to shut down ther'eactor is neglected in the large break analysis. l
- b. Injection of borated water provides for heat transfer from the core and prevents excessive clad temperaturet.
L In the,present Westinghouse design, the large break single ' failure is the loss of one RHR (low head) pump and the small break single failure is the loss of {
.one ECCS train. This means that for a large break, credit could be taken for two high head charging pumps, two safety injection pumps, and one low head L
pump; for the small break, credit could be taken for one high head charging pump, one safety injection pump, and one low head pump. The following is a discussion of the modeling procedure for the Westinghouse minimum safeguards and the flow splitting from a break of an ECCS injection line. ) l i The current design for both small and large breaks assumes that at least one i l train is available for delivery of water to the RCS. This means that one pump l in each subsystem delivers to the primary loop. For a large break analysis, a high head centrifugal charging pump starts and delivers flow through the injection lines (one for each loop) with one branch injection line spilling to the containment backoressure. To minimize delivery to the reactor, the branch line chosen to spill is selected as the one with the minimum resistance. 1:20v:'D/Os248s-1: 1 f,_ ~ . . _ .,.. - _
j , , . i B/B-UFSAR When the one safety injection pump, and the one low head residual heat removal j, .' ' pump start, flow is delivered to the reactor coolant system throug'h the ' 10-inch accumulator lines. One.line, with minimum resistance, is spilling to the containment backpressure.- ' ThefollowingdiscussionofECCSminimumssafeguardsLisforbreakswithan equivalent diameter less than a 10-inch accumulator line. For the high head centrifugal charging pump, the branch lines are 1-1/2 inches in diameter. Therefore, all small breaks with equivalent diameters less than the 1-1/2 i inches will have a spilling line to RCS pressure,and this flow will be considered lost to the break. . In the case of a small break, less than the 10-inch accumulator line out but greater than the 1-1/2-inch branch injection line,- the charging pump will spill to the containment backpressure. - t' Therefore in the ECCS analyses done by Westinghouse, single failure is taken { into account, i.e., loss of an RHR pump for large break or loss of one SI train for small break, and the spilling of the minimum resistance injection
- line. A break in an. injection line is of the small break category. '
Description of Laroe Break LOCA Transient The sequence of events following a large break LOCA are presented in Figure 15.6-3. - Before the break occurs, the unit is in an equilibrium condition, i.e., the 9 heat generated in the core is being removed via the secondary system. During blowdown, heat from fission product decay, hot internals and the vessel continues to be transferred to the reactor coolant. At the beginning of the blowdown phase, the entire RCS contains subcooled liquid which transfers beat from the core by forced convection with some fully developed nucleate boiling. Thereafter, the core heat transfer is based on local conditions with transition boiling and forced convection to steam as the major heat transfer
/ mechanisms.
1:20v:1D/0:24:s-14
} i y ~.n' g 8/8-UFSAR o, W The heat transfer between the RCS an'd the secondary system may be in either direction depending on the relative temperatures. In the case of continued j
' heat' addition to the secondary, secondary system pressure increases and the j
. main steam safety valves may actuate to limit the pressure. Makeup water to '
the secondary side is automatically provided by the auxiliary feedwater system. The-safety injection signal actuates a feedwater isolation signal ' which isolates normal feedwater flow by closing the main feedwater flow by I starting the auxiliary feedwater pumps. The secondary flow aids in the raduction.of RCS pressure. 1 l When the RCS depressurizes to 600 psia, the accumulators begin to inject [ borated water into the reactor coolant loops. Since the loss of offsite power j is assumed, the rea: tor coolant pumps are assumed to trip at the inception of
- l. 1 the accident. The effects of pump coastdown are included in the blowdown analysis. !
The blowdown phase of the transient ends when the PCS pressure (initially assumed at 2280 psia) falls to a value approaching that of the containment atmosphere. Prict to or at the end of the blowdou , the mechanisms that are responsible for the bypassing of emergency core cooling water injected into i the RCS are calculated not to be effective. Atthistime(called , end-of-bypass) refill of the reactor vessel lower plenum begins. Refill is - complete when emergency core cooling water has filled the lower plenum of the reactor vessel which is bounded by the bottom of the fuel rods (called bottom of core recovery time). The reflood phase of the transient is defined as the time period lasting from the end-of-refill until the reactor vessel has been filled with water to the )
' extent that the cors' temperature rise has been terminated. From the later y stage of blowdown and then the beginning-of-reflood, the safety injection accumulator tanks rapidly discharge borated cooling water into the RCS, contributing to the filling of the reactor vessel downcomer. The downcomer water elevation head provides the driving force required for the reflooding of the reactor core. The low head and high head safety injection pumps aid in l
the filling of the downcomer and subsequently supply water to maintain a full downcomer and complete the reflooding process. ) I 1:20v:10/0:24ss-1: ! I l
' B/8 UFSAR D , .
Continued operation of tho'ECCS pumps supplies water during long ters
. cooling.. Core temperatures have b a n reduced to long term steady state levels associated with dissipation of residual heat generation. After the water L
level:of the r6 fueling water storage tank reaches a minimum allowable value, cociant for long term cooling of the core is obtained by switching to the cold
-leg recirculation phase of operation in which spilled borated water is drawn from the engineered safety features sumps by the low head safety injection (residual heat removal) pumps and returned to the RCS cold legs. The containment spray system continues to operate to further reduce containment '
L pressure. Approximately 18 hours after initiation of the LOCA the ECCS is
' realigned to supply water to the RCS hot legs in order to control the boric acid voncentration in the reactor vessel.
Description of Small Break ~LOCA Transient ' As contrasted with the large break, the blowdown phase of the small break occurs over a longer time period. Thus, for the small break LOCA there are L only three characteristic stages, i.e., a gradual blowdown in which the decrease in water level is checked, core recovery, and'1ong term recirculation. l' , 1 15.6.5.3 Core and System Performance L 15.6.5.3.1. Nathematical Model The requirements of an acceptable ECCS evaluation model are presented in [ Appendix K of 10 CFR 50 (2). , Large Break LOCA Evaluation Model The analysis of a large break LOCA transient is divided into three phases: (1) b' lowdown, (2) refill, and (3) reflood. There are three distinct transients analyzed in each phase, including the thermal-hydraulic transient in the RCS, the pressure and temperature transient within the containment. and the fuel and clad temperature transient of the hottest fuel rod in the cere. Based on these considerations, a system of interrelated computer codes has been developed for the analysis of the LOCA. 1:20v:1o/Os24st-1s
~. .,__,,,.,.__._._.m. .._ __ _ . . _ . _ .. _ _ _.- _ __ _ . _ _ _ _ _ _
~ . . _ . _ _ _ . _ _ . _ _ _ _ _ _ _ _ _
- 8/B-UFSAR .
. l
' The~ description of the various aspects of the LOCA analysis methodology is given in' References 4, 10, 15, 20, and 21. l These documents describe the major phenomena modeled, the interfaces' among the computer codes, and ,
of the codes which ensure compliance with the Acceptance Criteria. The. ; SATAN-V!(Reference 5),WREFLOOD(Reference 6),COC0(Reference 7),BART (Reference 20), BASH (Reference 21) and LOCBART (Reference 8, 21)' codes are ! used to assess the core heat transfer geometry and to determine if the core remains amenable to cooling throughout and subsequent to the blowdown, refill and reflood phases of the LOCA. . SATAN-VI is used to calculate the RCS pressure, enthalpy, density, and the ' mass and energy flow rates in the RCS, as well as steam generator energy ' transfer between the primary and secondary systems as a function of time during the blowdown phase of the LOCA. SATAN-VI also calculates the i accumulator water mass a3d internal pressure and the pipe break mass and i energy flow rates that are assumed to be vented to the containment during 1 blowdown. At the end of the blowdown phase, these data are transferred to the I WREFLOOD code.: Also at the end-of-blowdown, the mass and energy release rates { during blowdown are transferred to the C0C0 code for use in the determination i of the containment pressure response during this first phase of the LOCA. e-Additional SATAN-VI output data from the end-of-blowdown, including the core L inlet flow rate and enthalpy, the core pressure, and the core power decay transient, are input to the LOC 8 ART code. WREFLOOD, using input from the SATAN-VI code, calculates the time to bottom of core recovery (B0C), RCS conditions at BOC and mass and energy release from the break during the reflood phase of the LOCA. Sines the mass flow rate to the containment depends upon the core flooding rate and the local core L pressure, which is a function of the Containment backpressure, the WREFLOOD l and COCO codes are interactively linked. The BOC conoitions calculated by { WREFLOOD and the containment pressure transient calculated by COC0 are used as ( input to the BASH code. Data from both the SATAN-VI code and the WREFLOOD L code out to BOC are input to the LOCBART code which calculates core average conditions at BOC for use by the BASH code. l L 1 i l 1:20w1D/0524ss-17 l l
8/8-UFSAR f 8 ASH providesa more realistic thermal-hydraulic simulation of the rocctor core and RCS during the reflood phase of a large break LOCA. Instantaneous values of the accumulator conditions and safety injection flow at the time of xY
' . completion of lower plenum refill are provided to BASH by WREFLOOD.. Figure 5.6-6 illustrates'how BASH has been suostituted for WREFLOOD in calculating transient values of core inlet flow, enthalpy, and pressure for the detailed fuel rod model, LOC 8 ART. A more detailed description of the BASH code is '
availabletin Reference 21. The BASH code provides a much more sophisticated treatment of. steam / water flow phenomena in the reactor coolant system during core reflood.. A more dynamic interaction between core thermal-hydraulics and system behavior is expected, and experiments have shown this behavior. The
+
BART _ code has been coupled with a loop model to form the BASH code and BART provides:the entrainment rate for a given flooding rate. The loop model ' determines the loop flows and pressure drops in response to the calculated core exit flow determined by BART. The updated inlet flow is used by BART to calculate a new entrainment rate to be fed back to the loop code. This-process of transferring data between BART, the loop code and back to BART forms the calculational process for analyzing the reflood transient. This coupling of the BART code with a loop code produces a more dynamic flooding ; transient. which reflects the close coupling between core thermal-hydraulics and loop behavior. The cladding heat-up transient is calculated by LOC 8 ART which is a combination of the LOCTA code with BART. A more detailed description of the LOC 8 ART code can be found in Reference 8, 21. During reflood, the LOC 8 ART code provides a ' significant improvement in the prediction of fuel rod behavior. In LOCBART the empiric.a1 FLECHT correlation has been replaced by the BART code. BART L emeloys rigorous mechanistic models to generate heat transfer coefficients appropriste to the actual flow and heat trarisfer regimes experienced by the fuel rods. i. 1 20v:1o/0531st-1s
*-+,c .**e-ewe-m----., ..~.*.#,-,. .+. --
.o ,
B/B-UFSAR The analysis in this section was performed with the' upper head fluid temperature equal to the reactor coolant system cold. leg fluid temperature, achieved by increasing the upper head cooling flow (Reference 19). Modeling features necessary to account for the reactor barrel-baffle region and the i reactor fuel assembly thimbles were included in this analysis as presented in References 15, 20, and 21. L 7he impact of a no single failure assumption for the ECCS was examined by re-aialyzing the most limiting break with maximum p' ECCS flows as required by Refarence 22. L Small Break LOCA Evaluation Model u< j ! For loss-of-coolant accidents due to small. breaks less than 1 square foot, the NOTRUMP, Reference 11, computer code is used to calculate the transient .l i depre:surization of the RCS as well as to' describe the mass and enthalpy of flow through the break. The NOTRUNP computer code is a state-of-the-art one-dimensional general network code consisting of a number of advanced features. Among these features are the calculation of thermal non-equilibrium ! I in all fluid volumes, flow regine-dependent drift flux calculations with counter-current flooding limitations, mixture level tracking logic in multipla-stacked fluid nodes and regine-dependent heat transfer correlations, i The NOTRUMP small break LOCA emergency core cooling system (ECCS) evaluation model ws: developed to determine the RCS response to design basis small break LOCAs and to address the NRC concerns expressed in NUREG-0611. " Generic p . Evaluation of Feedwater Transients and Small Break Loss of-Coolant Accidents J L in Westinghouse-Designed Operating Plants". 1 In NOTRUMP, the RCS is nodalized into volumes intercor.aected by flowpaths. p The broken loop is modelled explicitly, with the intact loops lumped into a ! second loop. The transient behavior of the system is determined from the governing conservation equations of mass, energy, and momentum applied throughout the system. A detailed description of the NOTRUNP code is provided in References 11 and 12. 1:20v:1o/0531:s-1:
.. ~ _ _-.__ _ . _ . _ . _ _ _ _ _ _ _ _ ' B/8-UFSAR The use of.NOTRUNP in the analysis involves, among other t representation'of the reactor core as heated control volumes with an associated bubble rise model to permit a transient mixture height calculation.
- - The multinode capability of the program enables an explic '
L detailed spatial representation of various system components. In particular. it enables a proper calculation of the behavior of the loop se loss of-coolant accident. Clad thermal analyses are performed with the LOCTA-IV cod uses the RCS pressure, fuel rod power history, steam flow past the u part of the core and mixture height history from NOTRUNP hydraulic , calculations as input. L Figure 15.6-53 presents the hot rod power shape utilized to per break analysis presented here. This power shape was chosen because it - provides an appropriate distributien of power versus core height and a local power is maximized in the upper regions of the reactor core 12 feet). This power shape is skewed to the top of the core with the peak l local power occurring at the 9.5 foot core elevation. This'isforlimiting process small breaks. for the small break analysis because of the core As the core uncovers, the cladding in the upper
- elevation of the core heats up and is sensitive to the local power at elevation.
The cladding temperatures in the lower elevation of the core, below the two phase mixture height, remains low. The peak clad temperature occurs above 10 feet. , Schematic representation of the computer code interfaces are giv 15.6-5 and 15.6-6. The small break analysis was performed with the 1985 NOTRUNP ve l Westinghouse ECCS evaluation model (refer to References 8, 11, 12
,and13).
l l l 1:20v:10/093189-20
L" - 8/8-UFSAR i .. The.use of NOTRLMP in the analysis involveiis, among other thin representation of the reactor core as heated control volumes with an associated bubble rise model to permit a transient mixture height calculation. The multinode capability of the program enables an explicit a - detailed spatial representation of various system components. - In[ , ' it enables a proper calculation of the behavior of the loop seal loss-of-coolant accident. Clad thermal analyses are performed with the LOCTA-IV code ( uses the RCS pressure, fuel rod power history, steam flow past the un g part of the core and mixture height history from NOTRUMP hydraulic' v. 1 calculations as input. , 4 Figure 15.6-53 presents the hot' rod power shape utilized to perform break analysis presented here. This power shape was chosen because it provides an appropriate distribution of power versus core height and also localpowerismaximizedintheupperregionsofthereactorcore(9feetto 12 feet). . This power shape is skewed to the top of the core with the peak local power occurring at the 9.5 foot core elevation. l This is limiting for the small break analysis because of the core unco process for small breaks. , As the core uncovers, the cladding in the upper elevation of the core heats up and is sensitive to the local. power at that elevation. The cladding toeperatures in the lower elevation of the core. below the two phase mixture height, remains low. l The peak clad temperature occurs above 10 feet. { 1 Schematic representation of the computer code interfaces are given in i 15.6-5 and 15.6-6. - , The small break analysis was performed with the 1985 NOTRUMP version o Westinghouse ECCS evaluation model (refer to References 8, 11, 12, an I 1920v:1o/0:3189-20 N &
- - . - - - - , - . . - - , , a , , - , , - , , - . - - , , - , , - . . . - , - - - - , ~ * - - - - - - - , , , , - - - - . . . - . --..,--,-,,,_-,--,,,-------..---,-------n--,~...w----,~.~ -.e--- -
. . .. ~ .- - _ _ _ _ _ _. ~ ___. ___ __ _ . . . __
B/8-UFSAR e , 15.6.5.3.2 Inout Parameters and Initial Conditions '
. I Table 15,6-2 lists important input parameters and initial conditions used in the analysis. .
5
.A range'of reactor operating temperatures were analyzed in order to lustify I operation at 100% power between 600 to 619.3 Degrees-F in the RCS hot' legs and .
535.6 to 556.7 in the RCS cold legs respectively. A full spectrum was done at i the reduc 6d operating temperatures and the most limiting break was repeated at the higher. operating temperatures. Additionally, the most limiting of the RCS , operating conditions was reanalyzed assuming no single failure (maximum safeguards) in accordance with the methodology presented in references 21 and 22. p l t The basis used to select the numerical values that are input parameters to the~ anal.ysis'havebeenconservativelydeterminedfromsensitivitystudies(refer r to References 16,'17, 18, and 21). In addition, the requirements of Appendix K. l regarding specific model features were met by selecting models which provide a significant overall conservatism in the analysis. The assumptions made pertain to the conditions of the reactor and associated. safety system equipment at the time that the LOCA was hypothesized to occur and include such assumptions as the core peaking factors, core power shape, the containment pressure, and the performance of the ECCS. The performance of the ECCS pumps ' has been degraded by reducing the flow provided by these pumps an additional ' l L 5% applied uniformly across the operational pressure range of'each individual : ECCS pump. This has been done to provide for a greater wear allowance over the anti:ipated lifetime of the ECCS pumps. Decay heat generated throughout y the LOCA transient is also conservatively calculated. ' A specific requirement for LOCA analysis (10CFR50, Appendix K) is that a range of power distributions shall be studied and the distribution resulting in the most severe calculated consequences shall be used in the analysis. The analysis was performed with a chopped cosine power distribution. ' Reference (21) addendum I has presented generic studies demonstrating that the chopped cosine power shape results in the most severe calculated consequences. 1920v:10/Os24st-21
3 . 8/8-UFSAR s , 15.'6.5.3.3 Result: l Based' on the results'of the LOCA sensitivity studies (References 15,17, and
- 18) the limiting.large break was found to be the double ended cold leg l
guillotine (DECLG). Therefore, only the DECLG break is considered in the large break ECCS performance analysis. Calculations were performed for a range _of Noody break discharge coefficients. The results of these calculations are summarized in Tables 15.6-1 and 15.6-3. ' 1 The mass and energy release data for the break resulting in the highest calculated peak clad temperature are presented in Subsection 6.2.1.5. l Figures 15.6-6 through~15.6-33 present the parameters of principal interest i from the large break ECCS analyses. For all cases analyzed transients of the l following parameters'are presented. !' A
- a. , Hot spot clad touperature (Figures 15.6-7, 15.6-7A&B, 15.6-22 I 15.6-28).
L,, . l p I L
.R Cociant pressure in the reactor core (Figures 15.6-8, 15.6-8A&B, 15.6-23,15.6-29).
c. Water level in the core and downcomer during reflood (Figures 15.6-9, (. 15.6-9A&B,15.6-24,15.6-30) ' d. t Core reflooding rate, (Figures 15.6-10,15.6-10A&B,15.6-25,15.6-31). e. Thermal power during blowdown (Figures 15.6-11, 15.6-11A&B, 15.6-26, 1 15.6-32),and i f. l Containment Pressure (Figures 15.6-12,15.6-12A&B,15.6-27,15.6-33). ' k The containment pressure transient resulting from a LOCA is presented in Subsection 6.2.1.5. l 1920v:1D/0531s9-12 w e w+- m-- w e--- ,wp-- e w-e wr -r= ---t wvwvi - e w y<--w+re-e-v.4- s- inw-w w--=a& --- w e w r-- - -, v---- w w-
6/8-UFSAR th For the limiting bresk analyzed, the following additional transient parameters are presented: ' e. Core flow during blowdown (inlet and outlet) (Figure 15.6-13).
, b. Core heat transfer coefficients (Figure 15.6-14).
- c. Hot spot fluid touperature (Figure 15.6-15).
l d. Ness released to Containment during blowdown (Figure 15.6-16). o . I l
- e. Energy released to Containment during blowdown (Figure 15.6-17). !
i ! f. Fluid density in the hot assembly during blowdown (Figure 15.6-18). l
- g. Mass velocity during blowdown. (Figure 15.6-20) i
- h. Accumulator water flow rate during blowdown (Figure 15.6-19), and i
)
1. Pumped safety injection water flow rate during reflood (Figure ! 15.6-21). i The maximum cladding temperature calculated for a hypothetical double ended [ [ severance of the RCS cold leg piping was 1883'F which is less'than the l Acceptance Criteria limit of 2200'F specified by 10CFR50.56. This result was i
- calculated assuming a discharge coefficient of 0.6 for the break and with the !
RCS operating at a nominal hot leg temperature of 619.3'F. Analysis performed i assuming the RCS to be operating with a, reduced hot leg temperature of 600'F I were found to be less limiting than the results obtained when the RCS was i assumed to be operating with a het leg temperature of 619.3*F. Additionally. *' the analysis with a nominal het leg temperature of 619.3'F was repeated assuming no single failure within the ECCS or ESF and the results were less limiting than the results calculated assuming a single failure in the ECCS. 1:20v:1D/0:24st-13 m.. . _ . . __ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ . _ _ . . - . _ _. _ _ _ . . . . - _ _ 1
g/8-UFSAR i i The maximum calculated local metal-water was 2.14% which isi embrittlement limit of'175 specified in 10CFR50.46. The total core' wide j metal-water reactions is less than 0.3% for.all breaks, as compared with the i 15 criterien of 10CFR50.46 and in all cases the cladding temperature transient I was terminated at a time when the core geometry.was st111' amenable to cooling. As a result, the core temperature will continue to drop and the ability to remove decay heat generated in the fusi for an extended period of time will be provided. These results provide assurance that operation with VANTAGE 5 fuel and with the RCS hot leg temperature in the range of 600 to 61g.3'F can be accomplished within the requirements of 10CFR50.46 and Appendix K to'10CFR50.46. Small Break Resulta l This section presents the results of a spectrum of small break sizes analyzed j for the Byron /Braidwood Stations. As noted previously, the calculated peak clad temperature resulting from a small break LOCA is less than that s calculated for a large break. Based on the results of LOCA sensitivity studies (Refarence 14 and 21) the limiting small brask was found to be less than a 10-inch diameter rupture of the RCS cold leg. The worst breaks size 1 (small break) is a 3-inch diameter break in the cold leg. This limiting break ; size was also analyzed for the reduced RCS operating tosperatures to show that ! the reduced temperature results in a less severe transient. The time sequence 1 of events and the results for all the breaks analyzed is shown in Tables 15.6-1 and 15.6-4. l During the earlier part of the small break transient, the effect of the break i flow is not strong enough to overcome the flow maintained by the reactor coolant pumps through the core as they are coasting down following reactor !
-trip.
Therefore, upward flow through the core is maintained. The resultant k heat transfer cools the fuel rods and cladding to very near the coolant ! temperature as long as the core remains covered by a two phase mixture. This 1 effect is evident in the accompanying figures. I
)
L l I 1:20v:10/0524as-24
t/8 UFSAR L L Figures 15.6-39 through 15.6-53 present the principal parameters of 4 for the san 11 break ECCS analyses. For all cases analyzed the follo! transient parameters are presented: a. RCS pressure (Figures 15.6-39A&4,15.6-46A&B,15.6-47) b. Core sixtur(height (Figures 15.6-40A&6, '15.6-48A&B,15.6-4g)
- c. '
- l. Not spot clad, temperature (Figures 15.6-41A&8,15.6-50,15.6-51),
E d. Core power after reactor trip (Figure 15.6-42) and o e. Pueped safety injection (Figure 15.6-52) For the limiting break analyzed, the following additional transient parameter are presented: l
- a. ' Core steam flow rate (Figure 15.6-43),
l b. l- Core heat transfer coefficient (Figure 15.6-44) and t c. Not spot fluid temperature (Figure 15.6-45). l The maximum calculated peak clad temperature for all small breaks analyzed; 1453.1*F. i These results are well below all acceptance criteria limits of 10CFR50.46 and in all cases are not limiting when compared to the results l presented for large breaks. ; i l 15.6.6 BWR Transient l This is not applicable to Dyron/Braidwood i 1 i l isaovaoms 4as-as l
8/s uFsAR ! 15.8.7 ggffgeg,nj n ,
- 1. '
' T. W. T..Burnett et al., 'LOFTRAN Code Description,' WCAP-7907, June .
1972. f Also supplementary information in letter fron T. N. Anderson. NS-TNA-1802, May 28, 1978 and NS-TNA 1824. June 18, 1978. - 2.
' Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors.' 10CFR50.48 and Appendia K of ICCFR50.
Federal Register, Volume 29, Number 3. January 4, 1974.
- 3. :
' Reactor Safety Study - An Assessment of Accident Risks in U. 5.
Commercial Nuclear Power Plants," WASH-1400, NUREG-75/014, October 19{
- 4. Bordelon, F. N., Massie. H. W. and Zordan T. A., ' Westinghouse ECC5 Evaluation Model - Summary,' WCAP 8339, July 1974. '
i 1 5. Bordelon, F. N., et al., '5ATAN-VI Program: Casprehensive Spacetime 'I Dependent Analysis of Loss of Coolant,' WCAP-8302 (Proprietary) and ! WCAP-8308 (Non-Proprietary), June 1974.
- 6. .
Kelly, R. D., et al., ' Calculational Model for Core Reflooding After a i Loss of Coolant Accident (WREFLOOD Code),' WCAP-8170 (Proprietary) and . WCAP-8170 (Proprietary) and WCAP-8171 (Non-Proprietary), June 1974.
- 7. Bordelon, F. N. and Murphy, E. T., " Containment Pressure Analysis Code (C0CO)," WCAP-8327 (Proprietary) and WCAP-826 (Non-Proprietary), June 1974.
- 8. Bordelon, F. N., et al., 'LOCTA-IV Progr e: Loss of Coolant Transient Analysis,' WCAP-8301 *
(Preprietary) and WCAP-8305 (Non-Proprietary), June 1974. .
- 9. Bordelon, F. N., et al., ' Westinghouse ECC3 Evalestion Model - '
Supplementary Information," WCAP-8471 (Proprietary) and WCAP-8472 (Non-Proprietary), April 1975. 1 sor.1D/os2449-se
8/8-UFSAR i l
- 10. Westinghouse ECC$ Evaluation Model - October 1975 Version,' WCAP-8622
*(Preprietary) and WCAP-8623 (Non-Proprietary), November 1975.
- 11. Weyer, P. E., 'NOTRUMP, A Nodal Transient Sas11 Break'and General Network Code,'WCAP-10079-P-A(Proprietary)andWCAP-10000-P-A(Non-Proprietary),
August 1985. l'
- 12. Lee, N.,
et al, 'We6tinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (Proprietary) and WCAP-10081-A (Non-Proprietary), August 1985. - )
- 13. Rupprecht, 5. D., et al, ' Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code,WCAP-11145-P-A (Proprietary),
l andWCAP-11372-A(Non-Proprietary), October 1986. 1 I
- 14. This Reference was intentionally deleted.
l
- 15. Johnson, W. J. and Thompson, C. M., ' Westinghouse Emergency Core Cooling
, System Evaluation Model - Modified October 1975 Version," WCAP-9168 (Proprietary) and WCAP-9169 (Non-Proprietary), September 1977, i
- 16. " Westinghouse ECCS Evaluation Model Sensitivity Studies,' WCAP-8341 (Proprietary) and WCAP-8342 (Non-Proprietary), July 1974.
p
- 17. Salvatori, R., " Westinghouse ECCS - Plan Sensitivity Studies.* WCAP-8340 (Proprietary) and WCAP-8356 (Non-Proprietary), July 1974.
l
- 18. Johnson, W. J., Massie. H. W. and Thoseson, C. M., " Westinghouse ECCS-Four Loop Plant (17x17) Sensitivity Studies,' WCAP-8566-A (Non-Proprietary), i July 1975.
1
- 19. Letter from T. W. Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Commission, letter number NS-TMA-2030, January 1979.
1:10v 1o/Ds34ss-37
S/8-UFSAR
- 20. Young, W. Y., 'lART-A1: I A Computer Code for the Best Estiaste Analysis of
{ tefleed Transients ' WCAP-9541-P-A.'Rev. I with Addendum 1-3 (Preprieta! March, 1984. ,
- 21. Kabadi, J. N. et al., 'The 1981 Version of the Westinghouse ECCS Evaluation Model using the BASH, Code 7. WCAP-10266-P-A Rev. 2 with Addenda (Preprietary) August 1986..
~ j 1
- 22. Letter ~from E. P. Rahe of Westinghouse Electric Corporation to Robert L. Tede of the U.S. Nuclear Regulatory Commission, letter number NS-EPR-2534 December 22. 1981.
l i b
+ i I
l l l ,. ) i 9 1:20v:10/Os24st-13
B/B UFSAR I i
. I
' TABLE 15.5-1 !
f!ME SEQUENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A - DECREASE IN REACTOR COOLANT INVENTORY ' i Tian Accident Event { (Sec) Inadvertent opening Safety valve opens fully 0.0 of a pressurizer t safety valve i low pressurizer pressure 41.7 i reactor trip setpoint , i i reached l Minimue DNBR occurs 42.6 i t Rods begin to drop 43.7 ! Large break LOCA i I i
- 1. Start DECLG CD = 0.8 0.0 i (TH0T=600.0'F) Reactor trip signal 0.505 l Safety injection signal 1.87 I Accumulator injection begins $1.1 1
l End-of-bypass . 23.47 ! End-of-blowdown 23.47 ! Pump injection begins 36.87 ! Bottom of core recovery 37.442 1 Accumulator, empty 58.46 4
- 2. Start i DECLG CD = 0.6 0'. 0 i (TH0T = 600.0'F) ,
Reactor trip signal 0.51 Safety injection signal 2.12 i i c
?
1:20v.10/ostost-ss
3/8-UFSAR ' l
. . i I
TABLE 15.5-1(Cont) i T!E SEQUENCE 0' EVENTS FOR INCIDENT 5 WHICH CAUSE A j s . DECREASE IN REACTOR COOLANT INVENTORY { Timo l Accident ivent (Sec) Accumulatorinjectionbegins 13.5 End-of-bypass { 31.077 i End-of-blowdown 31.077 l Pump injection begins 42.12 ; lottom of core recovery 46.02 i Accumulator empty 62.09 I
- 3. DECLG DC = 0.4 Start 0.0 (TH0T
- 600.0'F) Reactor trip signal 0.519 Safety injection signal 2.63 I Accumulator irijection begins 19.4 End-of-bypass l 40.584 End-of-blowdown 40.584 i Pump injection begins 42.63 l
Bottom of core recovery 56.986 ! i Accumulator empty 69.396 i
- 4. Start DECLG CD = 0.6 0.0 ;
(THOT*619.3'F) Reactor trip signal 0.597 i Safety injection signal 1.96 I Accumulator injection begir.: 14.2 End-of-bypass f 30.21 1 End-of-blowdown 30.21 Pump injection begins 41.96 i I I 1 itsor.10/os34:s-30 l 1
6/8-UFSAR
- TABLE 15.6-1(Cent)
TIME SEQUENCE OF EVENTS FOR INCIDENT 5 WHICH C DECREASE IN REACTOR COOLANT INVENTORY Time Accident , Event I (Sec) , Bottom of core recovery 44.007 Accumulator empty 61.65
- 5.
DECLE CD = 0.6 Start 0.0
.(TH0T*.619.3'F) Reactor Trip signal 0.597 Nazimum Safetyinjectionsignal 1.96 Safeguards Accumulater injection begins 14.2 (nd-of-bypass -
' M 21 End of-blowdown 30.21 Pump injection begins 41.96 Sottom of core recovery 44.597 Accumulator empty {
62.007 1 Small break LOCA i
- 1. 3 inch Start , 0.0 Nominal T,,, Reactor trip signal j 21.46 i
Safety injection signal 34.67 L Top of core uncovered 906.0 Accumulator injection begins N/A { Peak clad temperature occurs l ' 1592.0 Top of core covered 2689.9 '
- 2. 3 inch Start 0.0 Reduced T,y, Reactor trip signal 18.23 Safety Injection Signal 27.98 -
Top of core uncovered 996.0 Accumulator injection begins N/A Peak clad temperature occurs 1597.0 Top of core covered 2729.0 i 10sov:1o/082489-s1
, . , - . - . - - . - - . - - - - - - ' - - - ~ ' ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ .
i 8/8 UF5AR ' i TABLE 15.6-1(Cont) TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A ! DECREASE IN REACTOR COOLANT INVENTORY i
, Accident Time j Event (Sec) !,
- 3. 2 inch Start -
0.0 Nominal T Reactor trip signal avg 53.64 l Safety Injection signal 71.77 Top of core uncovered l N/A l Accumulator injection begins N/A < Peak clad temperature occura N/A i t Top of core covered N/A { I
- 4. 4 inch Start i
. 0.0 i Nominal T,y, bactor trip s 5 a1 12.74 Safety injection Signal 24.88 !
Top of core uncovered $24.0 .-
, Accumulator injection begins 879.02 .
t ' Peak clad temperature occurs 92).0 f Top of core covered 1618.0 i
- 5. 6 inch Start
' O.0 -
Nominal T,y, Reactor trip signal 8.25 { Safety injection Signal 14.41 ! Top of core uncovered 179.0 ; Accumulator injection begins 352.0 Peak clad temperature occurs 419.0 j Top of core covered 468.0 l 192ov:10/0:34st-ss t
,,.-.w-v,.- , ,~,-...,--,w,,,,,,,,,-v,_,,---,n~,__~ n-,,,--- , - , , , , , - - . , , - - - _
B/8-UFSAR i i TABLE 15.6-2 INPUT PARAMETERS USED IN THE ECCS ANALYS!$ Licensed Core PowerI , (Nwt) 3411 Peak Linear. Power, includes ,
, 1025 calorimetric factor (kW/ft) 13.8B I
Total Peaking Factor. Fg i 2.50 i Axial Peaking Factor, F Z 1.51515 ! t Power Shape t
'Large Break Chopped Cosine Small Break ;
See Figure 15.6-53 ~ l Fuel Assed ly Array : 17 x 17 VANTAGE 5 t i Accumulator Water Volume, nominal I ! (Ft/ Accumulator) 950 I I . Accumulator Tank. Volume, nominal ] I . t (Ft/ Accumulator) 1350 ! i Accumulator gas pressure, minimus (psia) 600 f Safety Injection Pumped Flow ; See Figures 15.6-26 and 15.6-52 i P Containment Parameters See Sect. 6.2 ' b t I Two percent is added to this power to account for calorimetric error. ' i 1 P J 1:30c10/983489-33
-- -- .,,,_..-.m..,._,....,,__...,,._,_,._.m_ ,. , -- .
8/8-UFSAR ! b ! i TABLE 15.6-2(Cont) l INPUT PARAMETERS USED IN THE ECCS ANALY3!$ j InitialLoopFlow(1b/sec) 9847, & 9569 1 l Vessel Inlet Temperature ( F) 535.6, & 556.7 Vessel Outlet Temperature ( F) 600.0, & 619.3 l L AverageReactorCoolantPressure(PSIA) 2200 Steam Generator Secondary Pressure (PSIA) 753, & 919 Steam Generator Tube Plugging Level (%) 15 i l l l It20v:1o/05244s-34
i B/8-UFSAR l o i i TABLE 15.6-3 ' LARGE BREAK LOCA RESULTS FUEL CLADDING DATA ! 1 Reduced That Reduced That Reduced That
-(600'F) (600'F) (600'F) g CD =0.8 {
.C0=0.6 Cg =0.4 j DECLE DECLG DECLG RESULTS FOR N LOOP )
1 l l Peak clad temperature (*F) 1823.64 1863.41 l 1527.01 $l Peak clad temperature ' l location l (Ft) 6.25 6.25 7.0 i Local Zr/H2O reaction maximum (5) 2.14 ' 2.96 0.546 l local Zr/H2O location (Ft) 6.0 6.0 7.0 l l I i Total Zr/H2O reaction (5) <0.3 <0.3 <0.3 Hot Rod burst time (Sec) 41.18 42.68 1 N8 1 Hot Rod burst location (Ft) 6.0 6.0 N/A l j I \ Burst was not calculated to occur for this case. NB "No Burst". l i l l l 1:30v:10/Os24:9-35 l
8/8-UFSAR i l TABLE 15.6-3 LARGE BREAK LOCA RESULTS FUEL CLADDING DATA l L Nominal That Nominal Thot (619.3'F) (619.3'F) MAX ECCS C0 =0.6 CD *0.6 DECLG , DECLG RESULTS FOR N LOOP Peak clad temperature ('F) 1983.07 i 1872.19 L Peak clad temperature Location (Ft) 6.25 6.25 Local Zr/H2O reaction Maximus (5) 3.26 3.15 .< g Local Zr/H2 O location (Ft) 6.0 6.0 ' l Total Zr/H2O reaction (%) <0.3 <0.3 i ! I I l Hot Rod burst time (Sec) 40.88 i 40.88, ' i Hot Rod burst location (Ft) 6.00 6.00 g i i s e 4 l l l 1910v:10/052489-34
B/B-UFSAR f TABLE 15.6 4 SMALL BREAK LOCA RESULT 5 FUEL CLADDING DATA i 2 Inch
- 4 Inch Results 6 Inch j Nominal T,,, Nominal T,,, Nominal T,,, i Peak clad temperature ('F) N/A 1293.5 1263.9 ,
Peak clad tagerature ; location (ft) N/A 11.5 11.25 ' Local Zr/H2O reaction, ! maximum i ( (%) N/A 0.10 0.07 ! i t l Local Zr/H 2 0 location (ft) N/A 11.5 11.25 l Total Zr/H20 reaction (%) N/A <0.3 <0.3 Hot Rod burst time (sec) N/A N/A N/A i. Hot rod burst location (ft) N/A N/A N/A
*There was no core uncovery for the 2-inch break.
9 6 192or.1o/0524st-37
- ~ . _ _ . . _ . _ _ _ . . . . _ _ _ _ _ . _ _ _ . _ - . . _ _ - - . . _ _ - . . _ _ . _ _ _ _ _ _ . _ . _ _ _ . . . - _ . - . - _ _ _ - - _ . - - . -
S/B-UFSAR i TABLE 15.6-4(Cont) i SMALL BREAK LOCA kESULTS FUEL CLA00!NG DATA -1 l
' l 3 Inch 3 Inch Results Nominal T,y, W uced T,,,
Peak clad temperature (*F) 1453.1 1424.5 Peak clad tosperatureP location (ft) 11,75 11.75 Local Ir/H 2O rea: tion. maxima (5) 0.48 0.42 Local Zr/H2 O location (ft) 11.75 11.75 i Total Zr/h20 reaction (5) <0.3 <0. 3. . Hot rod burst time (sec) N/A I N/A ' i
\
Hot rod burst location (ft) N/A N/A
. \
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BYRON /BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6 5 i Code Interface Description for Small Break Model <
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t i BYRON /BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT ! Figure 15.6 7 '
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l l ! I l 2 i BYRON /SRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT : Figure 15.6 7B ; Peak Clad Temperature - DECLG , (Lg=0.6) Minimus Safeguards Re5uced That (600.0 Deg F) ;
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SYRON/SRAIDWOOD STATIONS 1 UPDATED FINAL SAFETY ANALYSIS REPORT l i Figure 15.6 88 t Core Pressure DECLG 1
- l. (C,e=0.6). Minimum Safeguards j Re6uced Thot (600.0 Deg F) )
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BYRON /BRAIDWOOD STATIONS $ UPDATED FINAL SAFETY ANALYSIS REPORT t Figure 15,6 9 Reflood Transient Core & Downconer i Water Levels DECLG (Cd *0.8) Minimum Safeguares Nominal Thot (619.3 Deg F) '
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" BYRON /BRAIDWOOD STATIONS I UPDATED frINAL SAFETY ANALYSIS REPORT l Figure 15.6-9A Reflood Transient Core & Downconer Water Levels DECLG (Cg=0.8) 1 Maximum Safeguares l Nominal That (619.3 Deg-F) . . = . . . . _ _ _ _ _ .
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i I t 9 1 f e BYRON /BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6 98 Reflood Transient - Core & Downconer Water Levels DECLE (Cd *0.8) Minimus Safeguares Reduced That (600.0 Dog F) - y--d, e ee-+#s-w---+-*ew+v- , -.---e-,,,.m.,-www.-ve...w ,u .w-,-_m.. ..w..,e _-_m----..__-.--w-..
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I i t SYRON/SRAIDWOOD STATIONS .I ' UPDATED FINAL SAFETY ANALYSIS REPORT , Figure 15.6 10 i. I Reflood Transient Core Inlet Velocity i DECLG .8) Minimum Safe
. Noe Thot (819.3 DegF)guards
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TIME ISEC) i L ; BYRON /BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT L, Figure 15.6-108 Reflood Transient Core Inlet Velocity DECLG (C d *0.8) Minimum safeguards ReducM That (600.0 Deg F)
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I l SYRON/BRAZOWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6 11 1. L Core Power Transient 0.6) Minimum Safe DECLE NominI (Cg=l Thot (619.3 F) Deg guards \ L l
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1 i-BYRON /BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6-11A F- Core Power Transient I DECLG (C p0.8) Maximum Safeguards NominIl That (619.3 Deg-F)
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i L. TIME (SECONDS 1 t 1 l 1 l V' l 1 l BYRON /BRAIDWOOD STATIONS N' ' UPDATED FINAL SAFETY ANALYSIS REPORT p 7 Figure 15.5-12 Containment Pressure DECLG (Cec.6) Minimum safeguards NominIl Thot (619.3 Deg-F) t.
\ f' i _ ;_.2. _ .. . _ . _ _ _ _ _ _ _ _ - - - - - - - - --
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3 f l '.i 4 !v . i f w , ; BYRON /BRAIDWOOD STATIONS A UPDATED FINAL SAFETY' ANALYSIS REPORT i Figure 15.6-12A Containment Pressure L DECLG (Cro.6) Maximum safeguards NominIl Thot (619.3 Deg-F) t f
-.. + =m- - . , . . - - , - . . . - - +-...-...--....---..,..---v- - - - - -
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SYRON/SRAIDWOOD STATIONS ' UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6-128 Containment Pressure DECLG (C d *0.6) Minimum Safeguards ReducM That (600.0 Deg-F) m,
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. i BYRON /SRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6-13 Core Flow (Top and Botton DECLG (Cp0.8) Minimum Safegu)ards '
NominIl Thot (619.3 Deg-F) f
,.}+
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- 20. 40. 60. 80. 100. 120. 140. 160. 100. 200. 220.
TIME ISEC.) I 1 BYRON /BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6-14
- Core Heat Transfer coefficient DECLG (C d =0.6) Minieue Safeguards Nominal That (619.3 Deg-F)
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l i l BYRON /BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6-15 Fluid Temperature i DECLG (Cg=0.6) Minimum safeguards NominIl Thot (619.3 Deg-F)
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l l' l BYRON /BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6-17 Break Energy Released to containment DECLG (Cec.6) Minimum Safeguards NominIl Thot (619.3 Deg-F)
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1 l-l! 1, l-l l l i l l l BYRON /BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT I. Figure 15.6 19 i ) Accumulator Flow (Blowdown) DECLG (C =0.6) Minimum Saf uards
.Nomin 1 Thot (619.3 Deg- )
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L t i. SYRON/BRAIDWOOO STATIONS : UPDATED FINAL SAFETY ANALYSIS REPORT Figure 15.6-20 l Mass Velocity DECLG (C A .6) Minimum Safeguards NominIl Thot (619.3 Deg F)
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Figure 15.6-22 1-L Peak Clad Temperature - DECLG l' (Cg=0.8) Minimus Safeguards , Reduced Thot (600.0 Deg-F)
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l BYRON /SRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT l Figure 15.6 15 Reflood Transient Core Inlet Velocity i DECLG (Cg=0.8) Minimum Safeguards 1 Reduc 5d Thot (600.0 Deg F) l I .- .- . - . _ . . - . . - . , . .
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UPDATED FINAL SAFETY ANALYSIS REPORT l Figure 15.6-43 ! Core Exit Steam Flow (3-inch break) i Nominal Tavg = 588.0 deg F l l
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, Hot Spot Fluid Temperature (3 inch break)
Nominal Tavg = 588.0 deg F l l
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, UPDATED FINAL SAFETY ANALYSIS REPORT i Figure 15.6 53 l Wi!OreakPowerDistribution
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s,; . I I i ATTAC MENT B I LOCA RELATED TECHNICAL $PECIFICATIONS j Plant Name Byron /Bratewood Stations
< Larpe treak LOCA Analysis, Double toded Cold Leg 1
Gui lotine (DECLG) breaks, t =0.4,0.6,0.8, at the reduced hot leg temperat re of 600'F and i Cn=0.6 at the nueinal het leg temperature of i ST9.3*F. May, 1999. ) Small treak LOCA Ar.alysis, Cold Log Orifice Breaks, 2, 3, 4, 6, and 9 inch diameter breaks at the ; nominal hot leg temperatitre of 61g.3'F and the 3 inch break at th9 reduced hot leg tes9erature of i 600*F. May, IM9. ' t
)
Total Peaking Factor, FTg = 2.50 Enthalpy Rise Peaking Factor, F" g = 1.65 !
. . Cold Leg Accumulator l
t Water Volume: 3 g50 FT / accumulator (nominal) unchanged from previous analysis l Cold Leg Accumulator ! Gas Pressure: 600 Psia (Ninimum) unchinged from i previous analysis ) i K(z) Curve: Figure 1 No change to the K(z) curve as presented to Commonwealth Edison for the i Thot reduction program for Byron /8raidwood i l l
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1 Attachment E 1
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RA0!0 LOGICAL ASSESSMENi ' r !y !, FOR THE BYh0N/BAAIDMCOD STATION $ UNITS 1 AND 2 I
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, TRANSI'(ION TO WESTINGHOU5! 17X17 VANTAGE 5 FUEL AS5EMBLI M!
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RADIOLOGICAL A55ESSMENT Radialamical laures Tams far Accidents The extension of fuel burnup has been shown to have negligible impact on the core inventory of radioactive isotopes which are of concern in ! evaluating the radiological consequences of accidents (i.e., the short P half life noble gaser and iodines Thit has been documented Mth by i Westinghouse (Reference 1) and independent by). review perforssd for the NRC i (Reference 2). These reviewt c6nsidered burnuss of up to 40,000 M@/Mtv for the lead red which beund the fuel designs ning provided for Byrca arrj i Braidwood. [ Also of cancarn in evaluating the impact of extending fuel burnup on tM ~ t radiological consequences of accidents is the fraction of core activity j that is assumed to migrate out of the fuel matrix and into the fuel clad L gap region ( i.e. the fuel clad gap fraction) and thus available for .
.I release in the event the cladding is breached. The a:c1Jer.t analyses for Syrbh and traidwood have used the gap fractions roccensended by RMuluory Guides 1.25 trid 1.77. As indicated in f.eference 1, these gap fractions ;
remain extremely conservative for extended burnup fuel. Refei*ence 2 agrees that the extended burnup fuel gap fractions for most isotopes of : Guides, do not exceed the gap fractions specified in the Regulatory concern . t However. Reference 2 claims that for the Fuel Handling Accident (FHA) the i
!+131 gap fraction could be as high as 125 (an increase of 20% over the i Regulatory Guide 1.25 value). This increase is inconsistent with the !
i ; analysis performed by Westinghouse which modeled the anticipated fuel ! management and which determined a maximum gap fraction for 1 131'of just over one percent (Reference 1). i Fuel Handline Accident IFHA) . As discussed above, there would be no expected increase in the isotopic inventories or in the activity gap fractions. Thus the doses reported in the FSAR remain bounding. If the findings of Reference 2 are utilized, there is a 205 increase in r i the I 131 present in the fuel clad gap which translates into a 20% ! increase in the accident thyroid doses. Applying this increase to the FHA ! Site Boundary doses reported in the FSAR results in 24.4 ren for Byron (up i from 20.3 res) and 33 ren for Braidwood (up from 27.5 res). In each of ? these cases the dose remains well below the acceptance limit of 75 rom specified in th'e Standard Review Plan (Reference 3). t
I itana Canaratar Tuba tunture fitTR) i
' An evaluation was made to determine the impact of VANTAGE 5 fuel, 15%
Steen Generator tube plugging and a 3 percent reduction in the NSSV setpoint on the radiolog cal consequences of an SGTR event. The results i of the offsite dose eva untion show that both th 1 will decrease from those presented in the FSAR. yroid and whole body doses : Thus, the radiolojlical j consequences of an SSTR with VANTAGE 5 fuel are bounded by the ana ysis of ' record presented in the Byron /Braidwood FSAR. jtaneter taalant >= Locked Rotar (Lil ! The redislogical consequences of a LR event with VANTAGE 5 fuel are unchanged from those presented in r5AR Section 15.3. The results of the , FSAR analysis show that the thyroid and whole body doses at the site ! boundary and at the outer Doundcry of the It* pepulation zone are wtthin a . small fraction of the 10 CFR 100 Euidetina. ! Jgkad Reter with gggggggdggerfRtIfted Relief Valva (PORVMglgra _ l Thte doses for this event are also unchanged from thess presented in the .! FSAR. 100 guideline. Specifically, dotes will b7 within a small fraction of the 10 CFR { l \ \ l EEDIED
- 1.
WCAP 10125 P A, ' Extended Burnup Evaluation of Westinghouse Fuel *, - l December 1985. .' 2. NURES/CR 5008, ' Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors", February 1988. l
- 3. NUREG 0800 ;
Consequence,s of Fuel Handling Accidents', Rev. 1. July 1981.US N - 1 a t s v 9
4 , , . , , , , ATTACHMENT 6 ' r SIGN!flCANT HAZARDS EVALUATION
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.Significant' Hazards Evaluation n l' ,
for the Byron /Braidwood ! i
.o Transition to Hestinghouse i VANTAGE 5 fuel 1 1
c 1 I. HIR0DuCHON ; 1
- The Byron /Braidwood Stations, Units 1 & 2 are currently operating with Westinghouse 17x17 Optimized Fuel Assemblies (0FA). Commonwealth Edison ;
Company (CECO) has made the decision to replace the OFA fuel with Hestinghouse 17x17 VANTAGE 5 fuel assemblies. Implementation of this : decision, requires Commonwealth Edison to request an amendment to the ' Byron /Braldwood Facility Operating Licenses. This amendment. incorporates the changes to the Technical Specifications, presented in Attachment 2 of
.this app 1tcation, which are necessary to support a transition from a (Westinghouse) 0FA fueled core to a (Hestinghouse) VANTAGE 5 fueled. core.
The replacement of 0FA fuel with VANTAGE 5 fuel will be performed over a !'
' number of cycles. Initially the core will he comprised of both VANTAGE 5 and 0FA fuel. In subsequent cycles, more regions of VANTAGE 5 will be used until the core is entirely VANTAGE 5 fuel.
II DISIGN CHANGE ! The VANTAGE 5 fuel is a modification of the current OFA fuel assembly , design. The fuel assemblies are very similar to the current, previously
- approved 0FA assemblies. The VANTAGE 5 features are mechanically compatible with the current OFA fuel assemblier, control rods, and reactor internal interfaces. The VANTAGE 5 features are also hydraulically compatible with the OFA assemblies which will coexist ,
during the transition phase. A brief summary of the VANTAGE 5 design ; features and major advantages of the improved design are given below. i In_te.grALE.u.ellutnable_ Absorber _UIBAl - The IFBA features a thin boride coating on the fuel pellet surface on the central portion of the enriched uranium dioxide pellets. IFBA's provide power peaking and moderator temperature coefficient control. Illtelm_eA1 Ate _fl0ElilXeL.G.EH1 - Three IFH grids, located in the upper grid spans, provide an increased DNB margin. Be. cons.tltu. table _.lon_1tozzle_JRM1 - A mechanical disconnect feature facilitates the top nozzle removal. ILtendeL B.UInup_Capab11Lty - The VANTAGE 5 fuel design will be capable of achieving region average burnups in excess of 40,000 MHD/HTU. Changes in the design of both the top and bottom nozzles and increased fuel rod length increase burnup margins by providing : addltional plenum space and room for fuel rod growth. The basis for designing to extended burnups is contained in the approved Westinghouse topical HCAP-10125-P-A, " Extended Burnup Evaluation of Hestinghouse Fuel".
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.i e AttaLBlaukcis - The enriched fuel pellets in the top and bottom six inches of the fuel rod's active fuel region are replaced by natural 1 uranium pellets. This feature will reduce uranium requirements by 3 allowing lower feed enrichments, a lower number of feed assemblies or a combination of the two.
i In addition, both VANTAGE 5 fuel for Byron and Braidwood Stations, Units 1 and 2, employ the Debris F' Iter Bottom Nozzle (DFBN). The .
.DFBN contains a larger number of smaller flow holes as compared to
'the standard Westinghouse ~1ower nozzle. This design stops ;
, approximately.90%.of debris particles, those large enough to induce .
fuel: failures, while maintaining structural integrity and sufficient ! coolant flow. Implementation of the above VANTAGE 5 features will require.the following Technical Specification changes and revision to bases:
- Revised F (41) offset wings and gains with cycle specific identification d e to the VANTAGE 5 fuel design for OPDT/0 TDT set ,
- Added the HRB-2 DNB correlations, the Design and Safety Analysis DNBR limits and new FNDH values. These changes 'l reflect-the DNB correlations and the values for FNDH for the VANTAGE 5 and 0FA fuel. ;
- Revised rod drop time based on an increase in the hydraulic resistance arising from the VANTAGE 5 fuel design.
- Added new F(Q) limit, with cycle specific identification, F which reflects the value for F(Q) assumed in the safety .
analyses for the VANTAGE 5 fuel design. ; Revised the FNDH. j Revised bases discussions of DNB to reflect the new DNB correlations used for the VANTAGE 5 and 0FA fuel. l Changed the axial peaking factor multiplier for the F(Q) limit to reflect 9he value for F(Q) assumed in the s&fety analyses for either OFA or VANTAGE 5 fuel design.
- Revised Bases discussion for rod bow penalty reflecting the new DNB correlations used for the VANTAGE 5 and 0FA fuel.
- Revised the HTC Technical Specification to reflet.t increasing the MTC with burnup before decreasing the MIC n
towards EOL for the VANTAGE 5 core. Also added the statement, "Provlsions of Specification 3.0.4 are not applicable", to allow entry into Modes I and 2, if the requirements of the Action Statements are met.
- Replaced the 3 line segment K(z) Figure 3.2-2 with a 2 line segment curve to accommodate the VANTAGE 5 design.
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- k. , ' III Signlficant.llaZatd_Sumary w
e Based on the NRC approved Westinghouse document HCAP-10444-P-A, " VANTAGE
- l. 5 Fuel Assembly Reference Core Report", and Commonwealth Edison's review
,~ of VANTAGE 5 documentation provided by Hestinghouse in support of the ! VANTAGE 5 safety evaluation, it has.been determined that the changes associated with VANTAGE 5 fuel do not. involve a significant hazard. Additionally, a no significant hazards consideration was made because: a) J h e.s tc han g e s_d.o_Do Lintoly.e_Ls.i g rtiflc ant _in tre a s e_l a_t h e p tob ability _oLcoule Q u enceLof_an_ ac cide nt_prellou. sly _e yAl V ai ed . The Hestinghouse VANTAGE 5 reload fuel assemblies for the Byron /Braldocod Stations are mechanically compatible with the current 0FA fuel assemblies, control rods and reactor internals interfaces. The VANTAGE 5 ard 0FA fuel assemblies satisfy the safety criteria which form the current. design basis for the Byron /Braidwood Stations. Further, the reload VANTAGE 5 fuel assemblies are hydraulically compatible with the OFA fuel assemblies from the previous core. b). Ittls_shtostdneLant_tteate_thtonislhility_of_LneLor-. differ.ent kinLOLaccident_ftofL4Dy_ accid.enLDIellousiv enlUated. This is based on the fact that the method and manner of plant operation is unchanged. c) Ib es e_ chad ge Ldo..RQ.t_101ol1LLI.l g ulfic anL.te.du cilon..i n_a_ marg in_of.
.saf ely. In a mixed core of VANTAGE 5 and 0FA assemblies, the IFH grids in the VANTAGE 5 assemblies result in a localized flow i redistribution between adjacent VANTAGE 5 and 0FA assemblies. The affect of this localized flow redistribution is bounded by applying penalties to the transition core DNBR and large Break LOCA PCT results for those calculated in the analysis of a complete VANTAGE 5 fueled core. In addition, the core hydraulic resistance due to the IFM grids results in an increase in the control rod scram time to the dashpot from 2.4 to 2.7 seconds. This increase, as well as the other affects of the change in design, have been incorporated into the .
d non-LOCA and LOCA analyses, Indicating that the ANS Conditions II, III and IV acceptance criteria, endorsed by NRC HURE-0800, are still lj met, i In conclusion, the results of Commonwealth Edison's evaluations with i respect to the provisions of 10 CFR 50.92 demonstrate that the changes ! associated with VANTAGE 5 do not involve a significant hazard, j i i 1
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ATTACHMENT 7 4 f'c
. ENVIRONMENTAL ASSESSMENT p< , .>.
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, BRAIDH000 STATION UNIT 1 AND 2
' TRANSITION TO HESTINGHOUSE 17X17
. VANTAGE 5 FUEL '!
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s . ENVIRONMENTAL ASSESSMENT 4 ' p , L Commonwealth Edison has evaluated the proposed amendment against the criteria t for and.1 T **tcation of. licensing and regulatory actions requiring . environ u n C.sessr,ent in accordance with 10CFR51.21. It has been u determined that-t!e proposed change meets the criteria for categorical , exclusion as provided for under 10CFR51.22(c)(9). ; I Under this section a license 6 may qualify for categorical exclusion if: t Issuance of an amendment to a license for a reactor pursuant to Part 50 of 10CFR changes a requirement with respect to installation or use of a
- v. -facility component located within the restricted area, as. define in 10CFR h Part 20, or which changes an inspection or a surveillance requirement, 7 provided that (1) amendment involves no significant hazards L consideration -(11) there is no significant change in the types or significant increases in the amounts of any effluents that may be release off-site, and (111) there is no significant increase in individual or cumulative occupational radiation exposure.
Justification: L 1. The proposed amendment is being submitted pursuant to 10CFR50.90.
- 2. The proposed amendment involves an upgrade of the nuclear fuel which will be physically located in the Reactor or Fuel Handling Building. These two locations-are'inside the restricted area, as defined in 10CFR20.3.
- 3. The proposed amendment involves no significant hazards consideration as I determined pursuant to 10CFR50.92. The Significant Hazards evaluation is l
included in the licensing submittal. -1 l I
- 4. There'is no change in the types nor significant increase in the smount of any effluents that may be released off-site. The presence of or '
operations with VANTAGE 5 fuel will have no affect on the type or quantity of non-radiological effluent releases to an unrestricted area. l Based on design parameters of VANTAGE 5 and the operational experience of Callaway Station, Braidwood does not expect to see any change in the type ' of radioisotopes produced by operating with VANTAGE 5 fuel as opposed to I the OFA fuel currently in operation. Braidwood does anticipate some changes in the realistic source term, as defined in Byron /Braidwood UFSAR l- Chapter 11. Some changes expected would be an increase of tritium
.resulting from increase soluable 1:,oron concentration necessary to support L longer fuel cycles, and a decrease in fission products in the Reactor l Coolant due to improved fuel reliability. These changes are not anticipated to significantly increase the amount of any radioactive effluents released to an unrestricted area. In addition, all the l' radioactive effluent releases will be kept as low as reasonably achievable and in compliance with applicable 10CFR Part 20 limits.
1
- 5. There is no significant increase in individual or cumulative occupational radiation exposure. Based on design enhancements of VANTAGE 5 and I operational experience of Callaway Station with VANTACE 5, Braidwood does not anticipate any increase in individual or cumulative radiation exposure. All exposure will be kept as low as reasonably achievable and l:
in compliance with applicable 10CFR Part 20 limits. 1 1
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pe:. ,. . v .a; . ,,. o , I 1 e ENCLOSURE III p, ,
'l RESPONSES TO NRC STAFF QUESTIONS ON THE l '. '; ~
PROPOSED USE OF VANTAGE'S' FUEL- , 1 1 . 1 i
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lh' BESPONSESIO_NRGSIAEE QUE3 DONS _QlLTHEEBOROSED USE.OENAIGAGE5_EUELAllHERYHON45BAIDWOOQ_STADONS_UNIIS1AND.2 NBC-imoonadiknitations_orLUAm ot WCAP-104M
< . The NRC Staff reviewed Westinghouse's WCAP-10444, Reference Core Report VANTAGE 5 Fuel Assembly," and concluded in a staff Safety Evaluation Report (SER), Reference 1, that the generic topical report was an acceptable reference to support plant specific applications for use of VANTAGE 5 fuel, provided thideen conditions identified in the SER were addressed by the licensees. These thirteen conditions were appropriately considered in Commonwealth Edison's submittal for License Amendment Request, Reference 2. Each of the thirteen conditions is addressed below, and a reference is provided for the specific section in the Commonwealth Edison Licensing Amendment Request or other appropriate documentation where the condition is discussed.
NBC Goodition.L' The statistical convolution method described in WCAP-lD125 lbr the evaluation of initial fuel rod to nozzle growth has not been approved. This method should not be usedIn VANTAGE 5. Besponne; ' The statistical convolution method was not used for fuel evaluation. To determine the initial fuel rod to nozzle growth gap from fuel rod irradiation growth, the worse-case fabrication tolerances were used to evaluate fuel rod performance as discussed in Section 2.0, PaDe 8, of Attachment 1 to Reference 2. This evaluation was in compliance with Condition 1. Fuel rod performance for all fuel rod designs was shown to satisfy the Standard Review Plan fuel rod design bases on a region-by-region basis. NBC Gonditio1L2; For each plant appilcation, it must be demonstrated that the LOCA/selsmic loads considered In WCAP-9401 bound the plant in question; otherwise, additional analysis will be required to V demonstrate the fuel assembly structuralIntegrity. B98p0Dael The LOCA/selsmic loads considered in WCAP-9401 bound the Byron /Braidwood Stations Units 1 and 2. This is addressed in Section 2.0, Page 13, of Attachment 1 to Reference 2. < NBC.CGanditioo2: l l An irradiation demonstration program should be performed to provide early confirmation performance data far the VANTADE 5 design. Besponsel A demonstration program was successfully performed to determine early performance data on the VANTAGE 5 fuel assembly design features. The VANTAGE 5 demonstration program
- at commercial reactors is described in vection 1.0, Page 3 of Attachment 1 to Reference 2.
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, NRG.ConditionA For those plants using the ITDP, the restrictions enumerated in Section 4.1 of this report (SER) must ce addressed and information regarding measurement uncertainties must be -
provi%d. , Bespones; Westinghouse has addressed the restrictions enumerated in Section 4.1 of the NRC's generic VANTAGE 5 SER in a Westinghouse letter to the NRC Staff in March 1985, Reference 3. The ITDP instrument uncertainty methodology used for the Byron /Braidwood Units with VANTAGE 5 fuel was presented in WCAP-11656, Reference 4. Measurement ' L uncertainties were provided to the NRC Staff in Commonwealth Edison's letter, Reference 5. t1BC Condition 3: The WRB 2 correlation with a DNBR limit of 1.17 is acceptable for application to 17x17 ~ VANTAGE 5 fuel. Additional data and analysis are required when applied to 14x14 or 15x15 : l fuel with an appropriate DNBR limit. The appIlcability range of WRB-2 is specified in Section 4.2. . BespoDeel VANTAGE 517x17 fuel is proposed to be used at the Byron /Braidwood Stations Units 1 and
- 2. As described in Section 4.0, Page 17 of Attachment 1 to Reference 2, the WRB 2 l correlation with a DNBR limit of 1.17 was used for the VANTAGE 5 fuel with the i NRC-approved ITDP methodology. The WRB-2 correlation is supmrted by the DNB test data contained in Apandix A to WCAP-10444-P-A, and was appled within its appreted range of applicability "or the Byron /Braldwood Units.
l NRC ConditionD For 14x14 and 15x15 VANTAGE 5 fuel designs, separate analyses will be required la j determine a transitional mixed core penalty. The mlved core penalty and plant-cpecific safety t
- .ma in to compensate for the penalty should be addressed in the plant Technical '
ation Basts. Be690n6.91 l As noted above in Condition 5, VANTAGE 517x17 fuel is proposed for the Byron /Braidwood Station Units 1 and 2. The Westinghouse transition core DNB methodology as applied to the Byron /Braidwood Units is discussed in Section 4.0 of Attachment 1 to Reference 2. The Commonwealth Edison submittal, Reference 2 incorporates the NRC-approved change to the generic VANTAGE 5 transition core effects, Reference 6. The transition core penalty is covered by the margin maintained between the design and safety analysis DNBR limits. The proposed changee (see Attachment 2 to Reference 2) to Technical Specification Section 2.1.1, Reactor Core Safety Limits Bases, address these margins.
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Plant-specific analysis should be performed to show that the DNBR limit will not be violated with the higher value of FAH. Hesponte: The core DNB methodology as applied to the Byron /Braidwood Stations Units 1 and 2 with VANTAGE 5 fuel is presented in Section 4.0, Page 17, of Attachment 1 to Reference 2. The Commonwealth Edison submittal, Reference 2, contains Byron /Braidwood plant specific analysis results in Section 5.0 of Attachment 1 which support the use of FAH of 1.65 during , the transition period and with a full core of VANTAGE 5 fuel. All safety criteria are met wit 1 l an FAH of 1.65 as demonstrated in Section 5.0. NBC. Condition 1: The plant specific safety analysis for the steam system piping failure event should be performed with the assumption ofloss of offsite powerif that is the most conservative case. ( Bespontel l The Byron /Braidwood plant specific analysis did not require a new analysis for the Main l Steam Piping Rupture event. The current analysis is the Byron /Braidwood UFSAR which L included both the with and without offsite power cases is still applicable. The DNB design ; l basis was confirmed for the fuel transition by evaluation as presented in Section 15.2.4 of Appendix B of Reference 2. , NBC Gondition]; With regard to thee RCS pump shaft seisure accident, the fuel failure criterion should be the , 95/95 DNBR limit. The mechanistic method mentionedin WCAP-10444 Is not acceptable. f Besponae;
. l The mechanistic method was not used for the RCS :) ump shaft seisure (locked rotor)
! accident addressed in Section 15.3.3 of Attachment 3 to Reference 2. Any rods which i L violated the 95/95 DNBR limit were assumed to fall. Westinghouse performed two separate l and distinct analyses for the locked rotor / shaft break event. The first analysis, using the LOFTRAN and FACTRAN computer codes, was conducted to determine the wak RCS pressure, the peak clad temperature, and the amount of zinc-water reaction. T1ese results are presented in Table 15.3.2 of the VANTAGE 5 RTSR and were not used in any way to , determine the number of rods-in-DNB. 1 A second ana!ysis, using the LOFTRAN, FACTRAN, and THINC computer codes, was performed to determine the number of rods that experience DNB during the accident. The results of this analysis, which included increase in the Fo and FAH, showed that 0% of the rods were predicted to be below the 95/95 DNBR limit.' To reiterate, the mechanistic approach Identified in Reference 9 (NUREG-0562) of WCAP 10444 was not used in the Byron /Braidwood VANTAGE 5 License Amendment Request. l /sc1:0339T:3
NBC Gondition10: If a positive MTC is intended for VANTAGE 5, the same positive MTC consistent with the
. plant Technical Specification should be used In the plant-specific sakty analysis.
Bosponae; i Commonwealth Edison does not have any current plans to incorporate a positive MTC in the ' operation of the Byron /Braidwood Stations Units 1 and 2. Therefore, this condition is not
; currently applicable.
NBC Condition 2 The LOCA analysis rformeo for the reference plant with higher Fa of 2.55 has shown that the PCTlimits of 2 'F is violated during transitional mixed core, Plant specific LOCA analysis must be done to show that with the appropriate value of Fa the 2200*F criteria can be met during use of transitionalmixed core. Bespon8A i in accordance with Condition 11 of the VANTAGE 5 NRC SER, Byron /Braidwood Station Units 1 and 2 specific LOCA analyses were performed with consideration of transitional core effects. The Large Break LOCA analysis is summarized in Section 5.2.1 of Attachment 1 to Reference 2, ant detailed results are also provided in Section 15.6.5 of Attachment 4 to Reference 2. As described therein, the ECCS acceptance criteria of 2200'F is met for the l Byron /Braidwood Station with a LOCA Fn of 2.50. The worst case peak clad temperature is ! 1933.1F;it includes a conservative trans'Rion core penalty of 50*F. NRC CQDdillon.12; : Our SER on Westinghouse's extended bumup topical repon WCAP-10125 is not yet completel the approval of the VANTAGE 5 deslyn ibr operation to extended bumqq Ievels Is contingent on NRC approval of WCAP-10125. dowever, VANTAGE 5 fuel may be used to those bumups to which Westinghouse fuelis presently operating. Our review of the Westinghouse extended bumm topical report has not identified any safety Issues with
. operation to the bumup value given in the extended bumup report.
Besponse: WCAP 10125 has been approved (see Reference 7). The extended burnup methodology contained in this topical has been applied and is addressed in Section 2.0, page 8, of Attachment 1 to Referance 2. NBC Condition 13: Recently a vibration problem has been reported in a French reactor having 14 foot fuel assemblies; vibration below the fuel assemblies in the lower portion of the reactor vesselis damaging the movable Incore Instrumentation probe thlmbles. The staff is currently evaluating the implications of this problem to other cores having 14-foot long fuel bundle assemblies. Any limitations to the 14-foot core design resulting from the statievaluation must be addressed in plant specific evaluations.
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j l aesponae: The Byron /Braidwood Stations Units 1 and 2 have 12-foot long fuel assembly bundles and ) I therefore the above condition is not applicable. I o Beforence; i
- 1. to Westinghouse (E.P. Rabe) Re: "Acce>tance for i NRC Staff of Referencing letter. (C.O.
Licensing Thomas)l Report, WCAP-10444, ' VANTAGE Topica 5 Fue Assembly," ]
- undated.
- 2. NRC), Re:
- Application for Letter from RA Amendment Chrzanowski to Facility O)erating(CECO)
LicensestoNPF T.E.37Murley an (d NPF-77, NRC Docket Nos. i 50-454 and 50-455, datec July 31,1989 and letter from S.C. Hunsador (CECO) to T.E.
.Murley (NRC), Re: Apalication for Amendment to Facility Operating Licenses NPF 72 -i and NPF-77, NRC Doc (et Nos. 50-456 and 50-457, dated October 19,1989. j
- 3. Westinghouse letter, E.P. Rahe Jr. to C.O. Thomas t l
for additional Information on WCAP-10444 entitled (NRC), Response
" VANTAGE to Request No.1 5 Fuel Assembly" ;
E (Proprietary) NS NRC_85-3014, dated March 1,1985. ;
- 4. " Westinghouse improved Thermal Design Procedure Instrument Uncertainty Methodology," WCAP-11656, December 1987.
l 5. Letter from Docket and 2 ITDP," T. Tramm (CECO) Nos. 50-454, 50-455,to H.R. Denton 50-456 an (NRC) d 50-457, May 1982." Byro l L 6. Westinghouse letter, W.J. Johnson to M.W. Hodges (NRC), NS-NRC-87-3208,
" VANTAGE 5 DNB Transition Core Effects," October 2,1987.
- 7. Extended Burnup Evaluation of Westinghouse Fuel" WCAP-10125 P-A, December 1985.
NBC_OnnceIn f Inconsistency of values in FAH (B2-1 and B2-2) and 3/4 2-8. Besponte The values in both Sections of the Technical Specifications are consistent. The difference can be explained by the application of a 4% measurement uncertainty applied to the factor 1,49 factor in the bases Section resulting In a factor of 1.55 in Section 3/4 2 8. { t
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