ML19256F214
| ML19256F214 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/15/1970 |
| From: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Schneider R INTERIOR, DEPT. OF, FISH & WILDLIFE SERVICE |
| References | |
| NUDOCS 7912170436 | |
| Download: ML19256F214 (1) | |
Text
.
4 lNDIANA & MICHIGAN ELECTRIC COMPANY P. o. Bo X 18 BOWLING GREEN ST ATioH N EW YORK, N. Y.10004 December ll,1979 AEP:NRC:00297A Donald C. Cook Nuclear Plant Unit 2 Docket No. 50-316 License No. DPR-74
Subject:
Additional Data for Technical Specification Changes Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Denton:
This letter and its attachments provide additional information in support of Technical Specification change requests for Unit 2 made in our letter AEP:NRC:00297 dated November 2,1979. The first change request involves the maximum average Reactor Coolant System temperature as it relates to the overtemperature and overpower A T trips and the DNB limits.
This change was designated " Change No.1" in Attachment A to the referenced letter and the affected Technical Specification pages were included in Attachment B to the referenced letter.
The second group of change requests involves power peaking limits and were simi-larly designated " Change Nos. 3, 4 and 5" in the referenced letter.
Attachment C of the referenced letter provided a summary of analysis results as a technical basis for Change No.1.
Attachment A to this letter contains a complete description of the FSAR analyses results which required re-examination due to this change. These reanalyses were performed in accordance with the methods described in Chapter 14 (Unit 2) of the FSAR and the results show compliance with established safety criteria. The data being supplied provide the technical basis for the review and approval of the requested change.
In the discussion of the second group of change requests containea in our previous letter, reference was made to WCAP-9566, "The Nuclear De-sign and Core Management of the Donald C. Cook Unit 2 Nuclear Power Plant 90(
f
. (&
791217 0%
'jl 08 281
Mr. Harold R. Denton, Director AEP:NRC:00297A Cyle 2."
To facilitate your review of these change requests, this report is included as Attachment B to this letter.
Very truly yours,
F. W n JED:em ohn E. Dolan Vice President cc:
(w/o Attachment B)
R. C. Callen G. Charnoff R. S. Hunter R. W. Jurgensen D. V. Shaller -Bridgman 1608 282
Mr. Harold R. Denton, Director AEP:NRC:00297A bc:
(w/o attachment B)
S. J. Milioti/J. I. Castresana/P. K. Eapen/K. J. Vehstedt/V. P. Manno R. F.' Kroeger H. L. Sobel H. N. Scherer, Jr.
R. F. Hering/S. H. Steinhart/J. A. Kobyra J. F. Stietzel-Bridgman D. Wigginton -NRC Cook Plant NRC Resident Inspector DC-N-6015.1 AEP:NRC:00297 AEP:NRC:00297A 1608 283
.-r 2
ATTACHMENT A TO AEP:NRC:00297A 4
1608 284
TABLE 14.0-2 (Continued) 7 TIME SEOUE:?CE OF EVE'!TS Accident Event Time (Seconds)
Uncontrolled RCCA Bank Withdrawal at Power 1.
Case A Initiation of uncontrolled RCCA bank withdrawal at a high reactivity insertion rate (70 pcm/sec)
O Power range high neutron flux high trip point reached 1.8
~
Rods begin to fall into
'.j core 2.3 Minimus DN3R occurs 3.5
. 2.
Case B Initiation of uncontrolled RCCA bank withdrawal at a small reactivity insertion rate (2 pcm/sec) '
O Overtenperature AT react.or trip signal initiated 70.6 Rods begin to fall into core 72.6 Minicum D1:3?, occurs 73.2 UIIT 2 14.0-6 NC';D:EliT 1608 285
s
- TABLE 14.0-2 (Cor.tinued)
)
TDE SEOUENCE OF EVE';TS Accident Event Time (Seconds)
Rods begin to fall into core l'1 Shutdown =argin lost (if dilution continues after trip)
%5170 Loss of Forced Reactor Coolant Flow 1.
Four loops in op-All operating punps lose cration, four power and begin coasting 0
pumps coasting down
=
down Reactor coolant pu=p under-voltage trip point reached 0
Rods begin to drop 1.2 Minimun DMER occuss 2.7 2.
Four loops in op-eration, one pu=p coasting down Coastdown begins O
Low flow reactor trip 1.4 Rods begin to drop 2.4 Mini =us D:TER occurs 3.4 UNIT 2 14.0-8 A!EliE!E';T t
~
1608 286
TABLE 14.0-2 (Continued)
"b
- TIME SEQUE':CE OF EVE'.~rS Accident Event Time (Seconds)
Rods begin to drop 1.039 Maxi =um RCS pressure occurs 3.5 Maximum clad te=perature occurs 3.81 2.
Three loops in op-eration, one locked rotor Rotor on one pu=p locks 0
Low flow trip point reached 0.057 mb Rods begin to drop 1.057 Maximus RCS pressure occurs 3.30 Max 1=un clad te=perature occurs 3.81 Startup of an Inactive Reactor Coolant Loop Initiation of pu=p startup 0
Power reaches high nuclear flux trip 16.5 CIIT 2 14.0-10 AME:iDME'IT e
a-0 1608 287
TABLE 14.0-2'(Continued)
TIME SEOUENCE OF EVENTS Accident Event Tine (Seconds)
Rods begin to drop 17 ^
19.0 Minimum DNBR occurs Loss of External j
e Electrical Load 1.
With pressuriser control (BOL)
Loss of electrical load 0
10.4 Overte=perature aT 12.4 Reds begin to drop Minimum DIGR occurs
'14.0 4.s Peak pressuriser 14.0 pressure occurs 2.
With pressuri:cr control (EOL)
Loss of electrical" load,
O Overtenperature AT reactor trip point reached (1)
(1)
Reactor does not trip for this case.
UNIT 2 14.0-11 AvEmME;T 8
1608 288
TABLE 14.0-2 (Continued)
J TIME SEOUE'!CE OF E\\Ts7S Accident Event Time (Seconds)
Rods begin to drop (1)
Minimus DNBR occurs (2)
I Peak pressurizer pressure occurs 9
3.
Without pres-surizer control (ECL)
Loss of electrical load 0
High pressurizer pressure 5.1 reactor trip point recched Rods begin to drop 7.1 Minimus DNBR occurs (2)
Peak pressurizer pressure 9.5 occurs (1) Reactor does not trip for this case.
(T i
<.3R does not decrease below its initial value.
CIIT 2 14.0-12 AME:TD:fE:.7 h
O 1608 289
TIME SEOUEICE OF EVOITS Accident
' Event Ti=e (Seconds) 4.
Without pres-suri:cr control (EOL)
Loss of electrical load 0
High pressurizer pressure reactor trip point reached 5.1 Rods begin to drop 7.1 Minimun DN3R occurs (2)
Peak pressurizer pressurc occurs 8.5 Loss of Nor=al Feed-water and Loss of Off-site Pcwcr to the Station Auxiliaries (Station Blackout)
Lcw-low steam generator water level reactor trip; reactor coolant pumps begin to coast down 0
(2) DN3R does not decrease below its initial value.
UNIT 2 14.0-13 AEEfEIT 1608 290
TABLE 14.0-2 (Continued)
)
TIME SEQUENCE OF EVE';TS Accident Event Time (Seconds)
Rods begin to drop 2
Two steam generators begin to receive auxiliary feed from one cocor driven aux-iliary feedwater pu=p 60 Peak water level in pressurizer occurs 1340 Excessive Ecat Rc= oval due to m
Feedwater System Malfunctions one main feedvater control valve fails fully open 0
No reactor High neutron flux trip
'setpoint reached trip Minimum DNBR occurs 54.0 Excessive Load Increase 1.
Manual reactor control (30L) 10% step load increase 0
UNIT 2 14.0-14 A!CCME:iT
~.,
I s_s 1608 291 aw d
see g
TABLE 14.0-2 (Continued)
TIME SECUE!;CE OF EVC;TS Time (Seconds)
Accident Event Rupture of a Steam Line 0
1.
Case A Steam line ruptures 18 Criticality attained 14 Pressurizer e=pty 20,000 ppa boron 35.5 reaches core O
2.
Case B Steam line ruptures Criticality attained 13.5 18 Pressurizer e=pty 20,000 ppm boron 1
g*
reaches core 39.5 3.
Case C Steam line ruptures O
Criticality attained 18.5 14 Pressurizer e=pty, 20,000 ppa boron reaches core 36.5 4.
Case D Steam line ruptures O
Criticality attained 14 Pressuriner e=pty 18 20,000 ppm baron 39.5 reaches core
?
Uli1T 2 14.0-15 A! C ir: E ;T 1608 292
1.2 l
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=
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a 0.6 Hd 0.4
=
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~
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5 g
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.g 0.6 iS g
0.4
=
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2 4
6 S
10 TlHE (SEC0!OS)
UNIT 2 Figure 14.1.G-5 Nuclear Power and Heat Flux Transient, Loss of Power to Four Pumps with Four Locos Operating 1608 293
)
i N
2.6 2.4 OC m 2.2 o
2.0 1.8 0
2 4.
6 8
10 Tike.(SECONDS)
UNIT 2 Figure 14.1.6 6 DNBR Versus Time Less of Power to Four Pumps witn Four Lccas Operating 1608 294 e-e
1.2 l
l l
l
)
l.0 0.8 g
a' HOT CHANNEL O.6 g6
=
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0.8 2
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4 6
8 10 TlHE(SECONOS)
UNIT 2 Figure 14.1.6 7 Nuclear Power and Heat Flux Transients. Loss of Power 7
to One Pump witn Four Locr,s Ooerating t
.1.6 0. 8 2 9 5
- 12,906-5 2.8 2.6 E
2.4 E
3-J 2.2 2.0 0
2 4,.
6 8
10 TIME (SECONDS)
\\
UNIT 2 Figure 14.1.6-8 DNSR Versus Time Loss of Power ~ o One Pump with Four Loop! Operating 1608 296
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U E
NUCLEAR POWER
- j (FRACTIONOFHOMINAL)
CHANGE IN TAVG(f}
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10 20 30 40 50 UNIT 2 TlHE (SECONDS)
Figure 14.1.8 1 Loss of Load Accident with Pressurizer Soray and Power Operated Relief Valves. Beginning of Life g
l l
l l
)
y 580
?
5 560 WC
$ - 540 r"i 520
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0 10 20 30 40 50 TlHE (SEC0ttDS)
UNIT 2 Figure 14.1.3 2 Loss of Load Accident with Dres urizer Scray and Power Operated Relief Valves, Seginnir:g of Life.
O 1608 299
12.906-9
~
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1900 m
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11. 0 3.5 xg 3.0 o
2.5 2.0 I
l l
l.5 0
10 20 30 40 50 UNIT 2 TIME (SEC_0 fids)
Figure 14.1.8 3 Loss of Load Accident with Pressurizer ScraY an(f Power Operated Relief Valves,.End of Life 1608 300
=*
l 2,906-l 0 600 3
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\\
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650 W
2000 1750 1500
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1250 C
1000 m-M 750 5
500 250 E
O 0
10 20 30 40 50 TlHE (SECONDS)
UNIT 2 Figure 14.1.8 4 Lo:s of Lead Accicent with Pressurizer Sprcy and Power Operated Rel;ef Valves, End cf Life D""D "g~T J ih e Ju 2.1 1608 301
12.906-11 6.0 l
l 5.5 5.0 4.5 E
4.0 zo 3.5 3.0 2.5 2.0 2600 m
E 2500 v>0 2400 E 7 2300
${2200 5
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1900 1.2
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- = c 0.8 2*
=$
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f 590 d
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$O 570
~
ceE es5 560 cm 550 0
10 20 30 40 50 TlHE(SECO.NDS)
UNIT 2 s,
Figure 14.1.8 6 Loss of Load Accident %Nthcut Pressurizer Scray and Power J
Operated Relief Vaives, Begoning of Life 1608 303
12'.906-13 6.0 I
I I
I 5.5 5.0 4.5 4.0 m
m 3.5 g
3.0 2.5 2.0
~
1.5 2600 g
2500 E
21100 mW 2300
~
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2000 m
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xg 0. 11 E
k-0.2 I
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10 20 30 UN T 2 TIME (SECONDS)
Figure 14.1.8 7 Loss of Lead Accident Without Pres:urizer Spray and Power Operated Relief Va!ves. (nd of Liic i608 304
=
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I I
I E
580
.N 560 90 wo r-540 U
520 500 g
600 Rg 590 t5 580 "n
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E 560 E8 550 2000 g
1750 g"_
1500 1250 gg d~
1000 a: w 55 750 en '
WB 500 o
250 l
l l
1 0
0 10 20 30 40 50 UNIT 2 TIME (SECONDS)
Figure 141.8 3 Loss of Load Accicent 't.'ithout Pressunzer Scroy and P0wer Operated Re!iet Valves, End of-Life 1608 505
12,906-15 2325
)
u g
5 2300
- cg 2275 W
2250
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- t E6 580 8~
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20 40 60 80 100 UNIT 2 TlHE (SECONDS) 1608 506
-I Figure 14.1.10-1 Feedwater Control Valve Malfunctico
100 l
I
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i d
90 n$
80 Ee 70 Si y
60 50 6110 620 n
600 o
e3 580
\\
560 I
l I
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540 0
50 100 T50 200 250 300 TlHE(SECONDS)
UNIT 2 Figure 14.1.11 1 Ten Percent Stec Load increa:c. Beginninicf Life, Manual Reactor Control 1608 507
12,906-17
~*
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0.90 0
50 100 150 200 250 300 UNIT 2 TIME (SEC0tiDS)
Figure 14.1.11 2 Ten Percent Stco Load increase. Ecginnire-of Life, Manual Reactor Control t
1608 308
~
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70 g
g 60 50 640 620 C
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600 E4 580 g
560 l
l l
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540 0
50 100
'150 2C0 250 300 TlHE(SECONDS)
UNIT 2 Figure 14.1.11 3 Ten Percent Sten Load increase Ena of Life, Manual Reactor Contrat 1608 309
2600 l
I
)
g 2500 a
0 2t100 En m 5 2300
%E
- - 2200 5
my 2l00 g
2000 3.00 2.75 2.50 E
2.25 xo 2.00 i.75 i.50
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m5 0.95 I
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O 50 100 150 200 250 300 UNIT 2 TlHE (SEC0? IDS)
Figure 14.1.11 4 Ten Percent S eo Lead increa:c, End-of-Life, 7
Manuai Reac:cr Control l
1608 310
12,906-20 IU j
i i
i I
90 d
?>
80 y
70 5
/
i 60 D
50 660 640 620 E
o 600 g
-x 580 560 540-0 50 100 150 200 250 300 Ti!E (SECONDS)
UNIT 2 1608 311 Figure 14.1.11 5 Ten Percent Step Load increa:e Beginninyof Life, Automatic Reac:or Cpntrol
12.906-21 2600 1
I I-wg 2500 my 2tt00
- c. n g}
2300
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2000 l
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t 5
1.025 1
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1.000 0
50 100 150 200 250 300 T!HE(SECOND3\\
608 312 UNIT 2
-)
Figure 14.1.11 6 Ten Percent Stec Lead increase, Beginn ng of Life,
./
Automatic Reactor Control
3 100 l
i 1
90 a
W y
80 N
70 g
M f
60 r
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)
620 600 O
y 580 s
560 I
I I
l 5t10 0
50 100
'I SO 200 250 300 TIME (SECONDS) 1608 313 UNIT 2 Figure 14.1117 Ten Percent Steo Load increase, End of Life.
Automatic Reac:ct Contro\\
12.906-23 2600 l
- l l
l I
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A E~
2200 m"
' O 2l00 2000 3.00 2.75 2.50 k 2.25 gg 2.00 1.75 1.50
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50 100 150 200 250 300 UNIT 2 TlHE(SECONDS)
Figure 14.1.11 8 Ten Percent Step Load !ncreate, End of-Life, Automatic Reactor Contr51 1608 314
12,906-24
. 600 g
g g
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FRESSL'RIZER EMPTIES '
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- 5C 2 &*
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. STEAM FLOW = 3140 L3/SEC 5
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Figure 14.2.5 4 Steam Line Breck Ocv.n:: ream of Ficw Measuring No: le with d
Saf2:y injec: ion and Cut: ice Power (Case a) 1608 315
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Figure 14.2.5 5 S ?
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w o' 525 PRESSURIZER ENPTIES
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Figure 14.2.5-6 Steam Line Break Ocwnstream of Flow Meaturing Nozzle with Safety injec: ion, without Ou:: ice Power (Case c)
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550 g
Nw 525 PRESSURIZER E4PTIES T
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