ML20077F728
| ML20077F728 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 07/29/1983 |
| From: | Bradley E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8308030015 | |
| Download: ML20077F728 (23) | |
Text
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PHILADELPHIA ELECTRIC COMPANY 2301 M ARKET STREET P.O. BOX 8699 PHILADELPHIA, PA.19101
", ^ " ", ", * " " ", ' " ' ' " *
(215 84140o0 a.......a
.o..s..
EUG EN E J. GR ADLEY assocaava emessma6 counset DON ALD BLANMEN RUDOLPH A. CHILLEnti E. C. MI R M H A LL T. H. M AM ER CORNELL
,AuL AuER. AC" Al[ M BM 3
assistant osmanAL revusek EDW ARD J. CULLEN J R.
THOM AS H. MILLER, J R.
IRENE A. McMEMM A assis7 ANT cow.sEL Mr.
A.
Schwencer, Chief Licensing Branch No. 2 Division of Licensing U.
S.
Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
Limerick Generating Station, Units 1 and 2 Core Performance Branch and Reactor Systems Branch Re f erence :
PECO and NRC Telecon Dated 7/15/83 File:
GOVT l-1 (NRC)
Dear Mr. Schwencer:
As discussed in the reference telecon the information contained on the attached draft FSAR page changes will be incorporated into the FSAR, exactly as it appears on the attachments, in the revision scheduled f or Augus t, 1983.
Sincerely, a
1 n
Eug-e Bradley
/
RJS/gra/53 Copy to:
See Attached Service List t
1.f f
j[h -
5, 33 8308030015 830729 L
roa ^oocx osoooasa A
PDR U
c_
,0 cc: Judge Lawrence Brenner (w/o enclosure)
Judge Richard F. Cole (w/o enclosure)
Judge Peter A. Morris (w/o enclosure)
Troy B. Conner, Jr., Esq.
(w/o enclosure)
Ann P. Hodgdon (w/o enclosure)
Mr. Frank R. Romano (w/o enclosure)
Mr. Robert L. Anthony (w/o enclosure)
Mr. Marvin I. Lewis (w/o enclosure)
Judith A. Dorsey, Esq.
(w/o enclosure)
Charles W. Elliott, Esq.
(w/o enclosure)
Jacqueline I. Ruttenberg (w/o enclosure)
Thomas Y. Au, Esq.
(w/o enclosure)
Mr. Thomas Gerusky (w/o enclosure)
Director, Pennsylvania Emergency Management Agency (w/o enclosure)
Mr. Steven P. Hershey (w/o enclosure)
Donald S. Bronstein, Esq.
(w/o enclosure)
Mr. Joseph H. White, III (w/o enclosure)
David Wersan, Esq.
(w/o enclosure)
Robert J. Sugarman, Esq.
(w/o enclosure)
Martha W. Bush, Esq.
(w/o enclosure)
Spence W. Perry, Esq.
(w/o enclosure)
Atomic Safety and Licensing Appeal Board (w/o enclosure)
Atomic Safety and Licensing Board Panel (w/o enclosure)
Docket and Service Section (w/o enclosure)
I l
i
~
DRAFh 15.4.2.2.2 System Operation The focal point of this transient is localized to a small portion of the core.
Therefore, although reactor controls and instrumentation are assumed to function normally, credit is taken only for the RBM system.
A discussion of the transient follows below.
While operating in the power range in a normal mode (;;;;;t --
et-e in S;;ti:n 15.0.2.2.
of operation, the reactor operator makes a procedural error and withdraws the highest worth control rod until the RBM system inhibits further withdrawal.
Under most normal operating conditions, no operator action is required since the transient which would occur would bs very mild.
Should the peak linear power design limits be exceeded, the nearest local power range monitor (LPRM) would detect this phenomenon and sound an alarm.
The operator must acknowledge this alarm and take appropriate action to rectify the situation.
If the rod withdrawal error is severe enough, the RBM system.
would sound alarms, at which time the operator would acknowledge the alarm and take corrective action.
Even for extremely severe conditions (i.e., for highly abnormal control rod patterns, operating conditions, and assuming that the operator ignores all alarms and warnings and continues to withdraw the control rod),
the RBM system will block further withdrawal of the control red before the fuel reaches the point of boiling transition or the 1%, plastic strain limit imposed on the cladding.
15.4.2.2.3 The Effect of Single Tailures and Operator Errors i
The effect of operator errors has been discussed above.
It was shown that operator errors (which initiated this transient) cannot impact the consequences of this transient due to the highly reliable RBM system.
See Section 15.9 for details.
15.4.2.3 Core and System Performance
- Insed h 15.4.2.3;l Nathemati 1 Mo 1
For s tr nsient, he re tivity ertiorFFNteny slow.
Th fore, it is a equate o assu e that t le core pas suff4ciept
.t)fne to e uilibrpfe (1.e, that oth the autron fluz and' heat yuzare, inphase).
Ma ing use of the p ove asseption[ this Q/
4 transipit is 91culatpd using/ulics epiputer program / The/programsteady coupled nuclyse the al hydra is d ' scribed in det 1 in Re'f 15.4 inc1 ded in'the cal ulation.
Allspatial[ effects I
1/
15.4-5 Rev. 22, 07/83
o DRAVu'
~
,Dh Zkvert@
Tht:r eeed is c4scriko'in sdk 3.2.2. /..f of Gesta.E CEefi.c4-3)
)%e eye /e f, a cpe4 spec <)Su exa/ysis wa.1penGiad 7he cmse exces of Sis &ansied see te4fsie/ mi/d 4'eNAcc 3
/ pea teeef not press secanece o;S k.59 z%rifik, nor vb/s%
of -dhe / % p/s.16h s7fwk //~M oa &< afaaiuseccan1P74e
//why red psfreas is parded ix Rgate is 9-t.
,9 sami,<
of & venhku in AcP2 and N&M62 as a fusedzu of vi%g 5dm/
sf s lepswes <ss is,o,esedes% 7as/e ic+-2 wa
- a. se$out of /o'7% 6he tsd /s sham 'rb dhcA st Mc feel, resa/hky in a 4 cP2 of e./4/ ad N4Nex of /t,2 kw/fe, i
i i
e 9
l
m 0000-P05
~
LGS FSAN
^
,i kTheprimafoutputfromthi code, in additio [to bthe basi nuclear rameters,, ires hevariationof/helin r heat' genera r rate (LitGR): t variation of Jhe minimu critical power atio (MCPR); the otal reactor p6wer; and t e variation of i
the ncore ipatruments uring the te sient.
Afetectorresponse uses,j;ne instrum. nt responses o predict the RBM action co specified ondition for he rod wit raWal error.
iu er t in eva'luating the/ '
lytical met ods and ass mptions used,d to provide,a'
{
- .1
{
he y his transi fareconsidpre coparequences of alistic, yet onservative ssessment.o the consequ es.
15.4.2.3.2 nput Param ters and Ini al Condition {. '
/,,
s ThenumberofpossibleRWEtransiehtsisextreme.)[1drgeduef'.
t'o
(( b the number of conte I rods and tho' wide range pf expGsures a6d
' power 1,e'vels.
In der to encorppass all of ;he possible mTs ', - '2 which/ould coney embly occur,/a limiting an)alysis. is defided to c prov'de a conservative assessdent of the 6nsequences.
7, s
/
/
I~
Th conserv tive assumpti s ares
~
/'
~y highest wer h'rror is a cont ~ nuous withdrawal of the he assumed e a.
rod at its 'aximum drive spe'ed.
atedconditioks[
\\-
The r9 ctor is presumed to be in its/Theco/reIsass b.
/
/
I most reactive state c.
and devoid of allAienon.
This e gures that ths amount of facess reactigity which must be controll.e3 by the f
/
vable controk rods is :naxim m.
/
Furthermore[itisassumed hat the o tor has fully /
d.
inserted tfie highest worth rod prior'pto its removal, nd t
selecte#,the remaining,rontrol rod' pattern so that thermer limits are approached ip/ e fuel bundles j the th vic'rfity of the rod,to be withdrawn (see Figure 5.4-1).
l I
should be emphas'ized that fhis control rod f
onfiguration would be higtyly abnorral, and Id only f
be achieved by feliberate operator action or/by numerous
(
cperator errors.
/
/
/
e.
The operator is assu ed to ignore all enings during the transi'ent.
{
f.
Of th our LPRN strings neares o the copt'rol rod /
}
being withdr(aw6, the two highest reading 4PRM during' the transient ar assumed to have/f ailed g.
One two instrument c annels'is assumed to be
/
bypassed and out of servies.
ffe A and C LPR cha:rbers
/
input to one channel, whild e B and D cha:;@ 4:4ffput 15.4-6
- s.,.,e f~
)
onsyis w
v I
the other, cha el with the gr t
ssumed to be,-
passe.
J The e
.degr pnservative,pssumptions indicated abp e provide a high
's.
e of assurance that the transient, s analyzed, bounds all j RWE that cou,1d possibly/ occur.
Table 15.4-2 prese ts the other paameterspsedinthe7analysisof
~!
is transient 5.4.2.
.2.1 RBM ystem Operatien imizes the clonsequences of an RWE by blo M ng
[
he BM system mi e motion of th'e control re,d before the sa,f ety limitsare j
t exceeded.
/
The RBM has three trip levels (rod w,ithdrawal permissive
[
s removed)./ The tripflevelsmaybe,adjustedapdarenominal)that y 8%
\\
of reactor power apart.
The highest trip J4 vel is set so-the safsty limit / s not exceeded.
The lo4er two trip ledels are intended to prp ide a warning' to the ope'rator.
Settipq's are p
106%/ 98% and/90% of initial, steady, state, operating power at 100% flow.,/The trip levels are automatically varipd with reactor coolant f The va f atipn)cw to protect against fpel damage at lofer flows.
s is set to assure that,no fuel damage,will occur at any in i fed coolant flow.
The o rt t
b a given control rod withdrada/pe a or may encoun er any num e (up to three) of' trip points depending on,the starting power of l
1 l.
The lower two points pay be h
passed up (re' set ) b y m a n u s'] operation of a pushbutton,.
T h e r e s e't permissive /is actuated /and indicated by a light),4nen the RBM /
reaches 2%' power less,/than the trip point.
Tne,-cperator should then assess his local power and'either reset,cf select a ner' rod.
j The highest (powerT trip point may not be r,eset.
/
Its
,/
15.4.
.3.3 /'
/
,/
- The e useq,uences of t'his transient f
re relatively I
Neither locali::ed nor gross occurrence of/ oiling transi+.7on,d.nor violation of the'1% plastic str,a^tn limit on thp / cladding, occur i
~ThTvarT5H66:'ili Dis 'MCPR ~sndM2GR,'*5Y T 01IcW 4! T1thdra(a1
~
-of-the tigt}est worth Todr js' presented m
@609tes-44-4-gf&
'iME-E3 -respectively.
The bundles present
^1n'Figuresa d.4-2 and f
15.4-3 represent the eryvelope of the,HCPR and the M!. EOR f or eacf.
two-foot interval dur,tng the transient.
Var iatier/ sin the total i
reactor power is o shown in these figures.
Although these figur'es show th ange in therpa'l limits f r/o 't' he f ully i'nserted
/j j
l to t/le fully 4 hdrawn position, the contro,1 rod is automatica))y/,
i blocked a eet, even under' the worst s,et cf assu rptions.
%e' variat4-on in the signal res onse of t channels j
de 4heVn W F~iijdrep 15.4-4 and 15.4
,hp/'twc indepenge'nt
.'cWith a set i n t 6 f _1 g t.-
1 the red is shown to tlock t5f
, resulting in erOM B
/
and MPLHGR of 14.B4 kWf t.
~ "'i v
/
15.4-7
('
,/
,/f
(
f 15.4.
.3.4 Consi ratio of Un etaint' s
,e TF con rvativ assu
- tions, hich a ure t t this ransie.t as b ion 1,Y.4.2.3(p y anal zed, ha e bee cons, vativ previo' ly disc ssed
/
in S 2.
/
15.4.2.4 Barrier Performance An evaluation of the barrier performance was not made for this transient, since this is a localized transient with very little increase in total core power is less than 5% pically, the change in the gress core characteristics.
Ty and the changes in pressure,are negligible.
15.4.2.5 Radioloolcal Consecuences An evaluation of the radiological consequences is not required for this transient, since no radioactive material is released from the fuel.
N
~~ 15.4.3 CONTROL ROD MALOPERATION (SYSTEM MALFUNCTION OR OPERATOR ERROR)
This transient is covered by the evaluations cited in Sections 15.4.1 and 15.4.2.
15.4.4 ABNORMAL STARTUP OF IDLE RECIRCULATION PUMP 15.4.4.1 Identification of Causes and Frecuency Classifi_ cation 15.4.4.1.1 Identification of Causes This action results directly from the operator's manual action to initiate pump operation.
It assumes that the rer,aining loop is already operating.
15.4.4.1.1.1 Normal Restart of Recirculation Pump at Power This transient is categorized as an incident of moderate frequency.
15.4.4.1.1.2 Abnormal Startup of Idle Recirculation Purp This transient is categorized as an incident of moderate frequency.
15.4.4.2 Sequence of Events and Systen Coeration 15.4.4.2.1 Sequence of Events Table 15.4-3 lists the sequence of events for Figure 15.4-6.
(
15.4-B
l.
LGS FSAR DRAFT TABLE 15.4-1 SEQUENCE OF EVENTS FOR CONTROL ROD WITHDRAWAL ERROR IN POWER RANGE TIME-SEC flik 0
Operator selects and withdraws the highest worth control rod ma 4 5.0 The LPRM system indicates excessive 1ccalized
~
peaking, bu.t coaming o's lyceed.
~ 15. 0 The RBM system indicates excersive localized peaking, but weming h igncred t
7 630.0 The RBM system initiates a rod block inhibiting further withdrawal
@L
~ 60.0 Operater re-inserts control rod to reduce core power level
~ 8 0.,0 Core stabilizes at M. con itions O
s INPU PARAMETERS AND TI CONDITIONS FO.
CONTROL ROD WITHDRAW TRANSIENT 7
\\
Reactor er, NWt 3293 Average core expos e, NWd/t 60
\\
Xeno state ne i
Av rage li at heat gen ation rate kW/ft 5.34 k
simum inear heat neration ray kW/ft 1.34 x 10+8 i
/
/
Loca) oh of maximum LHGR bund 1A
( 21-3.8 )
p inimum CPR 1 298 Location of inimum CPR bbndle (21-38)
Highest w rth control od (26-35)
Redvi[drawalspe8, in./see 3.6 Core coolant fl ra'te Ib
/hr 1 0 x 10+8 Co e coolant niet enths py, Btu /lb 5.261 x 10 a
/
Cc re ave ge steam v lume fracti 3.71 x 0-1 Re or coolant pr sure, ave ege, psia 1.03 x 10+8 Control rod pattern igure 15 4-1 RBM trip setpoi t, %
106%
l ggyst PQsur p
l l
DRAW a
TABLE 15.4-X ROD WITHDRAWAL ERROR
SUMMARY
Rod Block Set Point Rod Position ACPR HLHGR
(%)
(Feet Withdrawn)
(Kw/Ft) 104 3.5 0.101 15.3 105 4.0 0.122 15.9 106 4.0 0.122 15.9 107 4.5 0.141 16.2 108 5.5 0.178 16.3 0
O D
AW,
New TA0 w "
N
76e eastvettES 185 CONTROL RODS
,)
h
~
J+1 3
S 7
9 11 13 15 17 18 21 23 26 27 29
')
1 I
1 1
1.
1.
5 4
4 4
4 51 47 8
18 16 18 6
9 4
4 3
3 4
4 36 16 16 42 44 42 18' 1,6 11 36 4
R 0
0 5
4 13 31 18 tt 44 43 44 16 18 15 27 4
B 0
0 8
4 17 23 16 16 42 44 42 16 1s 19 4
4 8
8 4
4 21 15 6
16 14 18 6
2$
4 4
4 4
11 07 16 16 16
[
1 L
_J
=
02 06 10 14 18 22 26 30 M
42 46 50 54 58
'04 LIMITS BONDLE (2106) dth b$
R m ' # *
- 't/ /
f k
[ (,[-/
LI4lERICK GENERATING STATION UNITS 1 AND 2 i
FINAL EAFETY ANALY518 REPORT ROD PATTERN FOR ROD
(-
WITHDRAWAL ERROR ANALYSIS, ROD (26 35)
FIGURE 15.41
l Dn;3T FIGURE 15.4-1 LIMITING CONTROL ROD PATTERN FOR ROD WITHDRAWAL ERROR I'
1 3
5 7
9 13 13 15 17 19 23 23 25 27 29 J
i 1
59 I
36 7
3 55 6
6 6
6 5
51 36 40 40 40 36
~
7 47 p
6 10 10 10 10 6
7 9
43 36 40 40 40 40 40 36 11 39 6
10 10 10 10 10 10 6
13 35 40 40 36 40 36 40 40 15 31 6
10 10 0
0 10 10 6
17 2;
40 40 36 40 36 40 40 19 23 6
10 10
- 10 10 10 10 6
21 19 36 40 40 40 40 40 36 b
1 23 15 6
10 10 10 10 6
~
II 36 40 40 40 36 27 07 6
6 6
6 03 b
36 CORE CDORDINATES 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58
" NER) Ptwes h
JUUU-Fil
= -
=-
I
- -m 9 F J'
/
1.1
~
1.5
=
1.5 za w
-i e V
EUNDLE (25-34) 1.06 O
14 k
=
E POWER a
1.3
=
1 04
_5 I
=
1 1
1 33 2
1.2 g
N BUNDLE (2138)
N i
'*I
~
Buf.CLE (23 34) e
!w 1 01 1.0 m
8 e
I I
I I
8 os tv 0
2 4
6 ff DIST ANCE CONTROL ROD WTNDR AWN thi 2
FWAL SAFETY ANALYSTS REPORT VARtATION OF FRACTION OF RATED POWER & MCPR WITH THE DISTANCE OF ROD (26 35) WITHDRAWAL DURING CONTINUOUS ROD WITHDR AWAL ERROR FIGURE 15.42 I
UDUDaY1g e
e e
r DRAF" 17 W'/
/
I i
j g
is 5
/
f EUNDLE (21381
/
BUNDLE (27 36) l
/
e
/
S 14
/
' / = N * ",,,,,,,,,.--=== "
8
- -~
.um, a,2.,
13 BUNDLE (25-34)
I y
12 d
8s 3,
l i
1 1 ___
1 0
2 4
8 8
to 12 CISTANCE CONTROL ROD WITH0m Av.N ffd LH8ERICK GENER ATING $7ATION UN6TS 1 AND 2 l
l FINAL SAFETY ANALY8tt MEPOMT l
l (
HEAT GENERATION MATE WITH THE VARI ATION OF MAXIMUM LINEAR DISTANCE OF ROD (26 35) WITH.
DR AWAL DURING CONTINUOUS ROD WITHDR AWAL ERROR FIGURE 16.4 3 i.
.. ~
GOOO-P13 a
DRAFT
~
CHANNEL A+C (24.41)
(32,41) 124.33) -
(32,33) g 120.
h I
s 110
=
A = NCNE F AILED 3 = 32,41 C *24,41 105 0 -24.33 f = 24,33 AND 24 41 i
J l
_1 I
I 1
I t
i 1
gg 0
1 2
3 4
5 6
7 8
3 to 11 12 13 CONTROL ROD FEET WITMDR AW*J agues LIMERICK GENER ATING STATION UNITS 1 AND 2 FtNAL BAFETY ANALY8it REPORT ROO BLOCK MONITOR RESPONSE TO CONTROL ROD MOTION, CHANNEL A+C FIGURE 15.4-4
a s.1 H V r-5 4 --
DRAFT 130
_CHANNELB+D 1
124.41) f32.41) 12g 1 (32,33) 2 (24.33) h 120 ' -
=
l
,1,. -
=
C v
A 110 D
I A
- NONE FAILED 8 *32.41 105 C +24,41 D = 24,33 E = 24. 33 AND 24,41
._L i -.i 1 - - - - 1_1_.L _
_J L
1 i
o i
2 3
4 s
e 7
E io 1,
,2
,3 CONTROL ROD FEET w6THDRAWN El67E T#j f i dr o S c=
L28ERICK OENER ATING $TATION UNITS 1 AND 2 FINAL SAFETY ANALY8t$ REPORT ROD BLOCK MONITOR RESPONSE TO CONTROL ROD MOTION, CHANNEL 8+D FIGURE 15.4 5
I' 0000-P1B DRAFT
~-
also put in an incorrect location Third,..the misplaced bundles would have to be overlooked during the core verffication performed following initial core loading.
,15.4.7.3.2 Frequency Classification This accident occurs when a fuel bundle is loaded into the wrong location in the core.
It is assumed the bundle is misplaced in the worst possible location, and the plant is operated with the mislocated bundle.
This accident is categorized as an infrequent incident based upon the following data.
Expected frequency:
0.004 events / operating cycle The abov'e number is based upon past esperience.
The only misloading accidents that have occurred in the past were in reload cores where only two errors are necessary.
Therefore, the frequency of occurrence for initial cores is even lower since three errors must' occur concurrently.
~
15.4.7.2 Secuence of Events an'd System Coeration The postulated sequence of transients for the misplaced bundle accident (MBA) is presented in Table 15.4-5.
Fuel leading errors, undetected by incere instrumentation following fueling operations, may result in undetected reductions in therral margins during power operations.
No detection is assumed and, therefore, no corrective operator action or automatic protection system functioning occurs.
15.4.7.2.1 The Effect of Single Failures and Operator Errors This analysis already represents the worst case (i.e., operation of a misplaced bundle with three SAF or SOE) and there are no further operator errors which can make the accident results any It is felt that this section is not applicable to this worse.
accident.
Refer to Section 15.9 for further details.
15.4.7.3 Core and System Performanc_e This event is discussed in Section 5.2.5.4 of_GESTAR II (Ref.
15.4-3). i a summary or the input parameters Yor snis analysis as given in Table 15.4-6 and Figure 15.4 g Results of analyzing the worst fuel bundle loading error are reported in Table 15.4-4. As can be seen,'MCPR remains well above the point where boiling transition would be espected to occur, and the MLMGR does not esceed the 1% plastic strain limit
.Therefore, no fuel damage occurs as a result Rev. 22, 07/83 15.4-14
l.
0000-P20 s
LGS FSAR DRAF TABLE 15.4-6 2K?UT PARAMETEkS AND INITIAL CONDITIONS FOR
/
,. FUEI/ BUNDLE LOADING ERROR
/
r
! Power, % rated /
100 ',
2.
Flop rate DO 3
tir.ated PR operp ing limit
[
..MLHGR o rating Jimit, kWit 13.4 5.,Avera - core
- aposure, t
O.
I 6.
Lo" tion f minimum CP b
die 25-34/
7.
cat n'of monimum bundle
( 2 '-4 6 )
{
8.
Co el rod pattern Figure 15. -8 NOTE:
Core condi ns are assume o be normal f a hot, operatingicore at BOC.
f( & W E W ' W
'%EW WW p cHeD
\\
,e y -, -,,
-,---.,,-.m-
--..,,-..-a-
=
- i..
DRAPT TABLE 15.4-6 INITIAL CONDITIONS AND RESULTS OF FUEL BUNDLE LOADING ERROR
- Reactor Power, % rated 100 Core Flow, % rated 100 For Largest ACPR:
Core Exposure, MWD /ST 7810 Location of Error (15,15)
Minimum CPR with Fuel Loading Error 1.29 Minimum CPR Operating Limit.
1.24 Minimum CPR Safety Limit 1.06 For Largest AMLHGR:
Core Exposure, MRD/ST 5000 Location of Error (11,04)
Initial LHGR (Assumed at Operating Limit), KW/FT 13.4 LHGR with Fuel Loading Error, KW/FT 16.97
- Core conditions are assumed to be normal for a hot, operating core.
//
"Aga) 7M8LE l
D000-P21, DRAF~
mS,
TABL [
f
/
NISPLACED BUNDL ALYSIS A
/
/
,/
/
~
2.1 w/o enric bundle et locafion 25-34),Adjacentto/ highest f F idg LPRM, oplaced with natsral pr'anium,A.711 w/o enriched bundle from ocation (23-60) at sero expospre.
The four/ bundles e/ound thef PRM are then assumed to be p t on the CPR limit by j
j
[
t'he operator.
/
EST) MATED A> CPR NCPlVa)
MCPRca)
>CPR LIMIT LIW1T M.P.B.A.
AMfj3 1.1[88
-0.1412
/ -10.78
/
2.19 w/o enriched bdndle a location (21-461,' replaced with natural uranium, O'711 w/O enriched bundle,from location'(23-60),/
f at zero exposure./ The fp6r bundles surrounding the LPRM are the.n assumed to be put on the MLHGR limit by the operator.
i ESTIMATED
?
AW:,HGR
/
MLHGR8 ML GR*
/
M:,HGR LIMIT /
LIM 3T _
M'. P. B. A.
AtEEER t
/
t'
/
13.4 g
1'.52 11.32 I
(G.41 l
l 1
I
/
cs)
/
Instrumented location
/
l (a)
Non-instrumented location
(
\\
\\(
g Oe2ere N
~
TABW C
.ame s e-i,,,,,ms.
e
DDDD-P22 704 ASSEM9L158
[
/
IN CONTROL RODS
- 'I
^
1 l l 1 b 9
9 11
,43' 17 19 31
_1,3
/
.j;
_s "
s 7
/
,/
42 f
j
/.
/
e
/
-7 j/
12 12 j/
'"' }
47 12
, 2 3
/
9 j/
/
j/
30 38
[
44 44 42 43 11
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'w LIIAERWIC OENER ATING STAT 40N UIRTS1 AND 2 7[
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FINAL SA'FETY ANALYSIS REPORT (S
CRtTtCAL ROD PATTERN AND FUEL BUNDLE EXCHANGE LOCATKWS FOR MISPLACEDSUNDLE ACCIDENT (0.0 GWd/tl rouRa ts.4.s i
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n
r LGS FSAR TABLE 15.4-10 INCREMENTAL CONTROL ROD WORTHS USING BPWSCD CONTROL BANKED CONTROL INCREASE I
CORE ROD AT ROD DROPS IN CONDITION GROUP (a)
NOTCH (X, Y)
FROM-TO Keff
~
bop,1 S ce A 7
12 26-35 0->48
.004658 G1 thr 4
11 h,e s a BOC-1, Sequence,A
'e' 12 26-43 0->
8
.0 2518 1 throu,gh G4,M/D all
't ers at 0
/
\\
BOC-Sequency A 9
4 30 0->8
.0021 hrotf'gh. G4 W/D G,1 -
rou G
8 at 12
/g g 0 at 0 BOC; S,equence A
10 4
2 1
0->8
.0 2141 G1 th ugh G4 W/D O(
7 G5, t ough G8 at 12
[
G t0 E
A 'The following assumptions were made to ensure that h
(
rod worths were conservatively high for the BPWS:
a.
BOC b.
Hot startup c.
No xenon I or definition of rod groups, see Figures 4.3-27 and F
4.3-28.
The worst case for each rod group is given.
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DRAFT
/0 TABLE 15.4[
INCREMENT WORTH OF THE MOST REACTIVE ROD USING BPWS Control Banked Control Increase Core Rod At Rod Drops In K sub Condition Group Notch (I J)
From-To (eff)
Sequence B 8
12 (26,55) 04 48 0.0097 0 GWD/ST Groups 1-6 Withdrawn NOTE: The following assumptions were made 'to ensure that the rod worths were conservatively high for the banked Pos; hon V)ilh draad Sepe,ee a)
BOC b)
Hot Startup (3Pd ).
c)
No Xenon
" " g t? C E p
Ale
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