ML20005G528
| ML20005G528 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/12/1990 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20005G523 | List: |
| References | |
| NUDOCS 9001190284 | |
| Download: ML20005G528 (305) | |
Text
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ENCLOSURE 1 . j[' Pf , PROPOSED TECHNICAL SPECIFICATION CHANGE .j k j
, SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 L DOCKET NOS. 50-327 AND 50-328' (TVA-SQN-TS-89-25) '
LIST OF AFFECTED PAGES
- b. Unit 1
+
VII - 3/4 2-5 3/4 2-6 - L 3/4 2-7 3/4 2-7a 3 3/4 2-8 3/4 2-9 3/4 4-15a
- 3/4 5-1
- 3/4 5-3 l
3/4 5-4 3/4 5-8 L 3/4 6-6a 3/4 6-21 B3/4 2-1 ' l B3/4'5 l l Unit 2 l-l VII 3/4 2-4 , 3/4 2-5 , I 3/4 2-6 3/4 2-6a 3/4 2-7 3/4 4-20 3/4 5-1 3/4 5-3 3/4 5-4 3/4 5-8 3/4 6-6a 3/4 6-21 B3/4 2-1 B3/4 5-1 9001190284 900112 PDR ADOCK0500g'"7 gg P
4 .- - .
-INDEX L
LIMITING-CONDITI'ONS FOR'0PERATION AND SURVEILLANCE REQUIREMENTS SECTION~ PAGE ; 3/4.5 EMERGENCY-CORE COOLING SYSTEMS (ECCS) j i 3/4.5.1 ACCUMULATORS Cold Leg. Injection Accumulators........................... 3/4 5-1
- ppe, ;;d Inj;;ti;n e::r r!:t:r;......................
... 3/4 S 3 3/4.5.2 ECCS SUBSYSTEMS - T grea'ter than or equal to 350'F. . . . . . 3/4 5-5 avg 3/4.5.3' ECCS SUBSYSTEMS - T,yg:less than 350 F.................... 3/4 5-9 -
j ( 3/4.5.4 20"0S' !!!XCTIOll O'l0Th - O EL ETE D p3pggfHed uvivu,aussw .vu ,_ . .. conn......................................
, in T1L < < , s si On O .( ,,_ m
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' 89 '*'3/4.5.5' r,
REFUELING WATER STORAGE TANK.............................. 3/4 5-13 1 3/4.6 CONTAINMENT SYSTEMS i i 3/4.6.1 PRIMARY CONTAINMENT
- Containment Integrity..................................... 3/4 6-1
\
. Containment Leakage....................................... 3/4 6-2 'i l
Containment Air Locks..................................... 3/4 6-7 Internal Pressure......................................... 3/4 6-9 Air Temperature...........................................
, 3/4 6-10 Containment Vessel Structural Integrity................... 3/4 6-11 ,
Shield Building Structural Integrity...................... 3/4 6-12 Emergency Gas Treatment System (Cleanup Subsystem)........ 3/4 6-13 Containment Ventilation System............................ 3/4 6-15 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System.................................. 3/4 6-16
~' '
Lower Containment Vent Coo 1ers............................ 3/4 6-16b ! R120 L SEQUOYAH - UNIT 1 l VII Amendment No. 67,63 116 June 1, 1989 1
h., - h - t I POWER DISTRIBUTION LIMITS v [ b 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-FK p LIMITING CONDITION FOR OPERATION 3.2.2 F q(Z) shall be 1 by the following relationships: l La Fq (Z) 5'[ [K(Z)] for P > 0.5 lK99 FS (2) 5 [ [K(Z)] for P $ 0.5 !EN 0.5 -
? >
where P = THERMAL POWER RATED THERMAL POWER l l- and K(Z) is the function obtained from Figure 3.2-2 for a given ' core height location. APPLICABILITY- NODE 1 , ACTION: l With F (Z) exceeding its limit: ( a. 0 Reduce THERMAL POWER at least R for each Nn F (Z) exceeds the limit within 15 minutes and similarly reduce the PowVr Range Neutron-R23 Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of r
-K,) have been reduced at least 3 (in AT span) for each M Fq(Z) eRceeds the limit, i.
l b, Identify and correct the cause of the out of limit condition prior l to increasing THERMAL POWER; THERMAL POWER may then be increased e provided its limit.q F (Z) is demonstrated through incore mapping to be within l l SURVEILLANCE REQUIREMENTS
- 4.2.2.1 - The provisions of Specification 4.0.4 are not applicable.
R99 fR: ":;;; 3/' d- 7e C SEQUOYAH - UNIT 1 3/4 2-5 Amendment N 19 January 23, o1989, 95 __ _. _. . . _ . m-
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- c. ,
' POWER DISTRIBUTION LIMITS
, ' SURVEILLANCE d i d EMENTS (Continued) .i 4.2.2.2 limit by:Fq(z) shall be evaluated to determine if F q (Z) is within its
- a. Using-the movable incore detectors to obtain a power distribu- *:
tion map at any THERMAL POWER greater than 5% of RATED THERMAL POWER. ; R23 ;
' b.~
Increasing the measured FQ (z) component of the power distribution ~ map by 3 percent to account for manufacturing tolerances and further , increasing the value by 5% to account for measurement uncertainties.
- c. -Satisfying the following relationship:
-(2.527 gg9 M
FQ (z) < - W(z)
* "f*)for P > 0.5-M II Fq (z) 1 x 0. for P 5 0.5 whereF$(z)isthemeasuredF(z)increasedbytheallowancesfor o .,
manufacturing tolerances and measurement uncertainty, F limit is I theF' limit,K(z)is.giveninFigure3.2-2,Pistherklative 1 THERMkLPOWER,andW(z)isthecycledependentfunctionthat L accounts for power distribution transients encountered during -) L normal operation. This function is given in the Peaking Factor R23 L1 Limit Report as per Specification 6.9.1.14. N
- d. . Measuring Fq (z) according to the following schedule:
- 1. Upon achieving equilibrium conditions after exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER L at which Fq (z) was last determined,* or o
- 2. At least once per 31 effective full power days, whichever .
occurs first. l L
*During power escalation at the beginning of each cycle, power level may be i increased until a power level for extended operation has been achieved and a power distribution map obtained.
l __ . , , , _ R99
- ~ , . . . , , ...
SEQUOYAH - UNIT 1 3/4 2-6 Amendment No. 19, 95 January 23, 1989
~
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POWER DISTRIBUTION LIMITS' y1.
.- )
SURVEILLANCE REQUIREMENTS (Continued) e.; With measurements' indicating:
-l maximum - FM (z) over z K(z)
~
' has increased since the previous determinatin of F M(z) either of the following actions shall be taken: 0
- 1. Ff(z)shallbeincreasedby2percentoverthatspecifiedin q
4.2.2.2.c, or .
- 2. F H(2) shall be measured at least once per 7 effective full-q t
power days until 2 successive maps indicate that I maximum (*) is'not increasing. over z K(2)
- f. Withtherelationshipsspecified-in4.2.2.2.cabovenotbeing satisfied:
1; Calculate the percent qF (z) exceeds its limit by the following expression: , h C] maximum Fkz)xW(z)
)
over z n -1 Ix100 for P > 0.5 C70 x K(z) R99 2.32 = *
)
!f
'l maximum Fq ,(z) x W(z)-)
-1 ove 7; 5 100 for P < 0.5 p
-x K(z) -
R99 l N ' A
- 2. Either of.the following actions shall be taken:
\.
- a. Place the core in an equilibrium condition where the l'
limit in 4.2.2.2.c is satisfied. Power level may then be increased provided the AFD' limits of Figure 3.2-1 are reduced 1% AFD for each percentq F (z) exceeded its limit, or R23
- b. Comply with the requirements of Specification 3.2.2 for Fq (z) exceeding its limit by the percent calculated above.
row. .,,", c , . R99 SEQUOYAH - UNIT 1 3/4 2-7 Amendment No. 19, 95 January 23, 1989
+*
s 3c . \!V% ~' ~)! w. 'x. - a y, e m l( N j ;' 7. ' im ;.-- e .- .. . . L 1; ^ , .-POWER DISTRIBUTION LIMITS 1 : ~;
- w , ,
-SURVEILLANCE REQUIREMENTS (Continued)
'i .
..,= 1 p#The limit 1 be 2.15 ead of 2. until an a sis in c ormance 4
Lwith FR.50.46, g plant op ting conditi and show g that a mit- R99
' g }g.:=
.237 satisf theLrequir nts-of 10 CF . 46(b). h been co eted -
and submit to NRC.' -
- i:
':if ~ ~ c 4
4 s . E y. C SEQUOYAH - UNIT 1 3/4 2-7a Amendment No. 95 R9 January 23, 1989
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(Q .9 4 POWER DISTRI'BUTION LIMITS-SURVEILLANCE-REQUIREMENTS (Continued) g '. The 111mits specified in 4.2.2.2.c, 4.2.2.2.e, and 4.-2.2.2. f above are'not applicable-in the following core plane regions:-
- 1. . Lower core region 0-to 15 percentLinclusive.
- 2. Upper core region 85 to 100 percent inclusive.
4.2.2.3' When Fq(z)~is measured for reasons other than-meeting _the requirements-of Specification 4.2.2.2 an overall measured F (z) sh'all be obtained from a 9
. power dist ion mapland increased by 3 percent to account for manufacturing ;q tolerances further increased by 5 percent to account for measurement uncertainty.
R23' t i I-e December 23, 1982 SEQUOYAH - UNIT 1 3/4 2-8 Amendment No. 19
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8 10 12 14 16 CORE 4GHT (FT) I: L l'. FIGURE 3.2 2. K(2)- NORMALIZED F (2) o AS A FUNCTION OF CORE HEIGHT r l l . : MAR 251982 SEQUOYAH - UNIT 1 3/4 2-9 Amendment No. 12
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Fibute 3. 2 ~ 2- - -
. (K(t)-'NormallaeQ F (a) Am. A Function.of. Cnet ' Hel3M g
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3.32 - , l,.- A!n M alaht - E t 't 1 L 0.000 1.000
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8.000.. 3.000
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30.000 0.040 13.000 0.935 - mm -
. = '
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, e e.e a.o 4.s e.e e.e as.e an.e core S ight Lesiunty enkated and accepted forissue.
s. 12/ /M signature bafe Ne idnpooVAR -uNITI 3/4 2-9 Ts suswAL A
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--TABLE' 3. 4 $7 f
-REACTOR COOLANT' SYSTEM PRESSURE ISOLATION VALVES.
l
? . VALVE NUMBER FUNCTION 63-560 Accumulator Discharge 63-561 : Accumulator Discharge: .
63-562 Accumulator Discharge' 63-563 Accumulator Discharge y 63-622 Accumulator Discharge l+ 623- Accumulator l Discharge 63-624 Accumulator Discharge-63-625 - Accumulator Discharge. 63-551- Safety Injection (Cold Leg) 63-553 . Safety Injection (Cold Leg) .
.63-557- Safety Injection (Cold Leg) i' p"4 63-555 Safety Injection (Cold Leg) 63-632- Residual Heat Removal (Cold Leg) -
j 63-633 Residual Heat Removal (Cold Leg) p , 1
~63-634- Residual Heat Removal (Cold Leg)
L 63-635 Residual Heat Removal (Cold Leg) ! l 63-641 Residual Heat Removal / Safety - ! Injection.(Hot Leg) 63-644 Residual Heat Removal / Safety : Injection (Hot Leg) 63-558 Safety Injection (Hot Leg)- l 63-559 Safety Injection (Hot Leg). L 63-543 Safety Injection (Hot Leg) [ l. L 63-545 63-547 Safety Injection (Hot Leg) Safety Injection (Hot Leg) ( -i 63-549 Safety Injection (Hot Leg) ! u 63-640 Residual Heat Removal (Hot Leg) 4 63-643 Residual Heat Removal-(Hot Lea) 87-558 f er Head ection L. b'6g0 87-55 Upper He Injection l' - 0 Uppe ead Injecti ' (43 7-561 U r-Head Inje on 87-562 pper Head I ' ction - 87-5 Upper Hea njection ;
'7* Upper d Injection harging h er)
FCV er Head Inje ion (chargi R87 (header) / FCV-74-1* Residual Heat Removal FCV-74-2* Residual Heat Removal
*These valves do not have to be leak tested following manual or automatic actuation or flow through the valve.
SEQUOYAH - UNIT 1 3/4 4-15a Amendment No. 39, 83 - September 21, 1988
w --. - - - g , a - V a 6 ; b s. ' i f- :- 3/4.5 l EMERGENCY CORE COOLING SYSTEMS (ECCS)
, i 3/4.5.1 ACCUMULATORS
~~~~.-~~
e - ~ COLD LEG INJECTION ACCUMULATORS
- LIMITING CONDITION FOR OPERATION 3.5.1.1_ Each cold leg injection accumulator shall be OPERABLE with:
- a. The isolation valve open, b.
A contained borated wataeborated water volume of between and gallons of C
- c. Between and ppm of boro
- DIT* re.mora l) and '
L jvs+ificabr. has d. A nitrogen cover pressure of between(W]060 6 and psig. in T5 cMnge~ APPLICABILITY: H0 DES 1, 2 and 3.* 8 81 - 2. 0 ACTION: o ,
. a.
' With one cold leg injection accumulator inoperable, excep,t as a result o of a closed isolation valve, restore the inoperable accumulator to i OPERABLE status within one hour or be in at least HOT STAND the next 6 hours and in HOT SHUTDOWN within the following 6 hours, l'
b. With one cold leg, injection accumulator inoperable due to the isola-tion valve.being closed, either immediately open the isolation valve or thebe nextin12 HOT STANDBY within one hour and be in HOT SHUT 00WN hours. i J ' c. With one pressure or water level channel inoperable per accumulator, be in at least HOT STANDBY within the next 6 hours and R128 SHUT 00WN within the following 6 hours,
- d. #
i With more than one channel (pressure or water level) inoperable per [ accumulator, inoperable, immediately declare the affected accumulator (s) , t L , l L
- Pressurizer pressure above 1000 psig.
l .
#CycleActions 4 refueling c and outage.d are in effect until the restart of Unit 2 from the R128 U
\
-o SEQUOYAH - UNIT 1 3/4 5-1 Amendment No. 124 i August 11, 1989
7: y fsy $his; spec.iSCAkion. i3 d1Id4cd. W ; Cy e sse ,
, EMERGENCY CORE COOLING' SYSTEMS (ECCS) t
[ j UPPER HEAD INJECTI'ON" ACCUMULATORS , .hlg,14, - MI'ITING CDflDITION FOR OPERAT , N-
/ ,/ , + _. , '3. .21Eachupperhe 'njectionaccumu i tor system shall e OPERABLE with:
l
- a. The isol ion valves open, i b; The ter-filled accumu tor containing be een 1805 and 185 cubic-- j fe of borated water av.ing a concentra on of between 19 and 00 ppm of boron, d p y o c The nitrogen be ing accumulator p ssurized to betw n 1185 and 1285 psig.
PLICABILITY: MOD 1, 2 and 3.* ACTION:
- a. W the upper head
- jection accumulat system inoperabl , except s a result of a'c sed isolation val (s), restore the per head
- injection accumul or system to OPE BLE status within ne hour or be'in at least T STANDBY within e next 6 hours a in HOT SHUT 00WN within.the f owing 6 hours.
C+ b. With the perheadinjecti accumulator system inoperable due t the iso tion valve (s) bei g closed, either ,imfnediately open th ! isola on valve (s) or-be n HOT STANDBY wit #in one hour and b in l HOT HUT 00WN within th next 12 hours. SURVEIL NCE REQUIREMENTS / l
- / , /
1: . 1.2 Each upper ead injection acc ulator system shs be demonstrated OPERABLE:
- a. At I ast once per 12 h rs by:
. Verifying the ontained borated ter volume and ni ogen pressure in e accumulators, d
7 2. Veri fyi that each accumu tor isolation val is open.
- Pressurizer Pr sure above 1900 p 'g.
(e-r i SEQUOYAH - UNIT 1 3/4 5-3 1 l
Y f-l '
, Deig
, EMER CORE COOLING SYSTEMS )-
, EILLANCE REQUIREMENT ntinue'd) -
.r (
- b. ; At' .ast^once per 31 days a .within 6 hours after ea solution ume increase of. great .than or. equal to 1% of nk' volume by verifying the boron c entration of the soluti in the water-filled
- accumulator.
- c. At least onc er 18 months by:~ s
- 1. V fying that each accumul or isolation valve clos automa-ically when the water 1 el in' the water-filled cumulator is 92.0 + 2.6/-5.8 inche above'the tank vendor w ing line when go l corrected for the ss of cover gas. /
- 2. Verifying t the total dissolved ni ogen and air in the water-fil d accumulator is less t g 80-SCF per 1800 cubi
.. feet o water (equivalent to 5 0 pounds nitrogen p ,
_po s water).
~
. d. A east once per 5 years removing the membrane talled between i he water-filled and ni gen bearing accumulato and verifying l that the ' removed memb ne bursts at a differe ial pressure of 40 + 10 psi. li
/
Thm page in+enknally lef+ blard - \ l i i
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l l l 1
)
t l i: SEQUOYAH - UNIT 1 3/4 5-4 Amendment No. 2ff,86 October 14, 1988 i
(17 ; 41 . .
, ; ?.
e.= l
- EMERGENCY' CORE COOLING SYSTEMS (ECCS)' / 1
./
-l-(f" l
' SURVEILLANCE-RE0'UIREMENTS.(Continued)
)
i
- h. By performing.a flow balance test during shutdown following- 1 completion of modifications.to the ECCS subsystem-that alter the subsystem flow characteristics' and verifying the following flow
_ rates: 1.- For safety injection pump lines with a single pump' running:. a.. The sum of the injection line flow rates, excluding the highest flow rate is greater thanlor equ o gpm' Land l R54
. 443
- b. The total pump flow rate is less than or alto 675gpm.-[R54
- 2. - For centrifugal charging pump lines with a single pump running:
- a. The sum of the. injection line flow rates,' excluding.the
' highest flow rate is greater than or equa to gpm, and lR54- ,
309 1
- b. The total pump flow rate is less than o ual to 555 gpm.
- 3. For all four. cold leg injection lines with a single RHR pump gpm. <
running a flow rate greater than. or equal t@ Ct L l l- . I 1 i ( December 1, 1986 SEQUOYAH - UNIT 1 3/4 5-8 ' Amendment No. 50
,m - - , -
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? TABLE'3.6-1 (Continued)- .
g BYPASS' LEAKAGE PATHS TO THE AUXILIARY BUILDING. g
. SECONDARY' CONTAINMENT BYPASS LEAKAGE PATHS a-5 x ~
e PENETRATION. DESCRIPTION: -RELEASE LOCATION - V c-
- a X-106 Postaccident Sampling, Air Discharge. ' Auxiliary Area to Containment' X-108 -OHE AAaintenance Penetra+ son Auxiliary Area ~ ~c
, X-109 -bttt- Main +cn ance penc+ ration. Auxiliary Area ' .
-X 110 'J:lI %:i!inq fn; 'R9,4-f ,
X-114 Ice Condenser ' Auxiliary Area. , X-115 Ice Condenser ' Auxiliary Area- - Auxiliary Area: X-116A Postaccident Sampling, Containment - Air Sample w . m e O Ds a 0 3> Ok O3 (T C1 m3 V2 w? e
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e O
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s
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c, L _, ; w
.-ps, ,aw .g* g,g1 .s g - w'sg yy 9q +e hT MW W % 'd' D FN -'M'W
- T4ft'- +f% 'meit'visu' f m- t e'- w*.8 'Wt V '"'e' -% ' "'
1
, TABLE'3'.6-2 (Continued) -
g CONTAIMENT ISOLATION VALVES , k VALVE NUMBER FUNCTION MAXIMUM ISOLATION TIIE (Seconds) E A. -PHASE "A" ISOLATION (Cont.) 2 .2 [ 61. FCV-77-19 RCDT and PRT'to V H ' 10*
?! 62. FCV-77-20 N2 to RCDT .10*-
H 63 FCV-77-127 Floor Susp Pump Disch 10* w 64. FCV-77-128 Floor Susp Pump Disch 10*
- 65. FCV-81-12 Primary Wati. Makeup 10*
- 66. V-87-7 -
UH1 st Line -1 R82 '
- 67. FCV-87 Test Li
- Nd . FCV- -9 I Test L 10*
- 69. F 10 UNI Test ine~ 10*
- 70. CV-87-1 UHI Te Line 10*
B. PHASE "B" ISOLATION ' q 1. FCV-32-80 Control Air Supply 10 '
- 2. FCV-32-102 Control Air Supply -10 .- ;
y 3. FCV-32-110 Control Air Supply 10 N
" 4. FCV-67-83 ERCW - LWR Capt Cirs 60 -
- 5. FCV-67-87 ERCW - LWR Capt Cirs 60* g4g~
- 6. FCV-67-88 ERCW - LWR Capt Clrs 60* i
- 7. FCV-67-89** ERCW - LWR Coot Cirs 70* l R86
- 8. FCV-67-90** ERCW - LWR Capt Cirs 70* *
- 9. FCV-67-91 ERCW - LWR Capt Cirs 60*
- 10. FCV-67-95 ERCW - LWR Capt Cirs 60*
- 11. FCV-67-% ERCW - LWR Capt Cirs 60* R41 i
- 12. FCV-67-99 ERCW - LWR Capt Cirs 60*
g m 13. FCV-67-103 ERCW - LWR Capt Cirs 60* 3 4 14. FCV-67-104 ERCW - LWR'Capt Cirs 60* 5 ? 15. 16. FCV-67-105** FCV-67-106** ERCW - LWR Capt Clrs ERCW - LWR Capt Clrs 70* l R86 70* I EE 17. FCV-67-107 ERCW - LWR Capt Cirs 60* ge 18. FCV-67-111 ERCW - LWR Capt Clrs 60* .
,_ 19. FCV-67-112 ERCW - LWR Capt Cirs- 60* L!
w e 20. FCV-67-130 ERCW - Up Capt Cirs 60*
- E 21. FCV-67-131 ERCW - Up Capt Cirs 60 R41 ;
u
- 22. FCV-67-133 ERCW - Up Capt CIrs 60* '
- 23. FCV-67-134 -
ERCW - Up Capt Cirs 60* 'l
~
m 24. FCV-67-138 ERCW - Up Capt Cirs' 60*- m y w --
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b .. n .i f: .. L, ' i 0 ;3/4.2f POWER DISTRIBUTION LIMITS s - L , , BASES >
' The: specifications'of this section provide assurance of fuel integrity ,
L during' Condition I (Normal Operation) and II (Incidents of Moderate Frequency) >
' events'by: (a) maintaining the calculated DNBR in the core at or above design 6-7
-during normal operation ~and in short term transients, and (b) limiting the ,
fission gas' release, fuel pellet temperature and cladding mechanical properties to-within assumed design criteria. In addition, limiting the peak linear power density during Condition IE events provides assurance that the initial conditions assumed-for the LOCA analyses are net and-the ECCS acceptance criteria limit of.2200*F is not exceeded. The definitions of certain hot channel and peaking factors as used in these specifications are as follows:
. F (2) Heat Flux Hot Channel Factor, is defined as the maximum local 9
heat flux on the surface of a fuel rod at core elevation Z divided ' by.the average fuel rod heat flux, allowing for manufacturing: tolerances on fuel pellets and rods.
'FN Nuclear Enthalpy Rise Hot Channel Factor is defined as the ratio of the
'1Negral of linear power along the rod with the highest integrated ' power to the average rod power. '
13/4.2.1 AXIAL' FLUX DIFFERENCE (AFD) The'11m AXIAL FLUX DIFFERENCE assure that the F (Z) upper bound n envelope of times the normalized axial peaking factor is.not exceeded during'either normal operation or in the event of xenon redistribution follow-
-ing power changes.
Provisions for monitoring the AFD on an' automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the one minute average of each of the OPERABLE excore detector outputs 1 and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 !. of 3 OPERABLE excore channels-arm outside the allowed AI-Power operating space l and the THERMAL POWER is greater than 50 percent of RATED THERMAL POWER. I 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, RCS FLOWRATE AND o NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The l'imits on heat flux hot ' channel factor, RCS flowrate, and nuclear enthalpy rise hot channel factor ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event w, of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit. ' December 23, 1982
'SEQUOYAH - UNIT 1 8 3/4 2-1 Amendment No.19 I
e - e~.w..- 3 .-. _. - ,. ___,._.______,,,________________________.____.m_
;i The limits in t.he speciCeation for accumuistee vetime and nHrogen cover ~ pressure are analysis limi4s end do nvi includs instrument wcerkMy. The cover pressure i limiis were determined by Weslighouse to be GISpsic and 697. SpsiQ . Sincc the ins 4rumen+ read -ovis in the conical room a re in have beer 1 converted to psi 4 and rounded sot the ne,osig are3 >+wh the T3 yalves ole numben.
l, The a e+04l M Sf*Jen cover -)bressur6 ea limit 5 in Sq ya3 gg5jgn l
% documenis C **0 3 Psif *"d '82'8 psif.
^
3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS De lele. The OPERABILITY of each cold leg injectionbMrded3[js[.kfhd accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core in the event that the RCS pressure falls below the specified pressure of the accumulators. J g tpe y cold lea iniection accumulators, this condition occurs in the avant of a larneArupture.I T upp he in tio t ViotAoth yie It)/ge p6d 5%11 bjiak)fbe t[i g o rati WC The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met {AdcI INSERT &]
~
The accumulator power operated isolation valves are considered to be
' " operating bypasses" in the context of IEEE Std. 279-1971, which requires that
( bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required. The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required. 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long term core cooling capability in the recircu-lation mode during the accident recovery period. SEQUOYAH - UNIT 1 B 3/4 5-1 *
.4- 't: .
'!!N SsR T. :B
'}' .
- u. -
u Th'e ~ minimurn %ron: con'cenirubion en30res tha.6 the i-reac+or- core eiii remain sobcnWca / during tho Iacc orn vla +or inj eckon. period of a small brwh .
' LOCA,
L,v i2%} ; c r ' INDFX: I c LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS Q SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 -ACCUMULATORS
-ColdLegInjectionAccumulators........................... 3/4 5-1
'.5pr ;; . Ldni;vn L ; lau,e......................... 3/4 5 3 !
3/4.5.2 'ECCS SUBSYSTEMS T,yg greater than or equal to 350 F..... 3/4 5 l 3/4.5.3 -ECCS SUBSYSTEMS T ava less than 350'F.................... 3/4 5-9
- This.
cArnge- .3/4.5.4- "0;Gn wouiGn ;V;!O4 D E L E TE C) Was 2 $UbmiN E = ; I;i ;; ki' T; * - .- - 2lA 5 11
'l in D 1 c.hi 6 ";;t Tr; 6 ;..............................................
-3/4 0 12 '
-M- . -
3/4.5.5 REFUELING WATER STORAGE TANK.............................. 3/4 5-13 3/4.6' CONTAINMENT SYSTEMS "
'3/4.6.1' PRIMARY CONTAINMENT
+
Containment Integrity..................................... 3/4 6-1 i
~
a S Containment Leakage....................................... 3/4 6-2 Containment Air Locks..................................... 3/4 6-7 Internal Pressure......................................... 3/4 6-9 . L . Air Temperature........................................... 3/4 6-10 Containment Vessel Structural Integrity................... 3/4 6-11 Shield Building Structural Integrity...................... 3/4 6-12 Emergency Gas Treatment System (Cleanup Subsystem)........ 3/4 6-13 L Containment Ventilation System............................ 3/4 6-15 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System.................................. 3/4 6-16 1 Lower Containment Vent Coo 1ers............................ 3/4 6-16b R61, April 4, 1988 SEQUOYAH - UNIT 2 VII AmendmentNo.//,61
m pq ' 9
.' L
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M POWER DISTRIBUTION LIMITS-sg ' C 3/4.2.2' HEAT FLUX HOT' CHANNEL FACTOR-F q @- LIMITINGf0NDITIONFOROPERATION 3.2.2 F (Z) shall_be_ by the f611owing relationships: n ' R95 o
-F9 (Z):5 [K(Z)] for P > 0.5 R95 Fn (Z) 5 [+0.5-9 F ] [K(Z)] for P 1 0.5 THERMAL' POWER j i
'where P = RAILD THERMAL POWER !
~
+ and K(Z) is the' function obtained from Figure 3.2-2 for a given-4 - core height location.
T' W APPLICABILITY: MODE.1 ACTION: . With F (Z) ' exceeding its- limit: I 9
- a. Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds-the limit 9
within 15 minutes and similarly reduce the Power Range Neutron . Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION ;
'may proceed for up to a total of 72 hours; subsequent POWER OPERATION 3
-may proceed provided the Overpower Delta T Trip Setpoints (value of K4 ) have.been. reduced at least 1% (in AT span) for each 1% F (Z) 9 ,
~ exceeds the limit.
- b. -Identify'and correct the cause of the out of limit condition prior
, to increasing -THERMAL POWER; THERMAL- POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within
- 9 E
its limit. SURVEILLANCE REQUIREMENTS L 4.2./.1 The provisions of Specification 4.0.4 are not applicable, f: # Ses Pew s ;<1 2 Ce R95 L ? L SEQUOYAH - UNIT 2 3/4 2-4 Amendment No. 21, 95 March 10, 1989 h
l [4- ,
~$
~
POWER DISTRIBUTION LIMITS
; SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.2 F q(z)'shall be evaluated to determine if gF (Z) is within its limit by'
.. . . . i
~ a- Using the movable incore detectors to obtain-a power distribution' l map at any THERMAL' POWER greater than 5% of RATED THERMAL' POWER. <
- b. Increasing the measured F (z) component of the power distribution R21 9
map by 3 percent to' account for manufacturing tolerances and further - increasing the value by 5% to account for measurement uncertainties. 4 c. Satisfying:the following relationship: I
. R95 F (z) 5 . ., x K(z) for P > 0.5 ,
.P x W(z)..
R95 F (z) 3 2. R;- x K(z) for P s 0.5
.W(z) x 0.5 where F (z) is measured Fq (z) increased by the allowances for manufacturing tolerances and measurement uncertainty, Fq. limit is the-Fq limit, K(z)'is given in Figure 3.2-2, P is the relative l THERMAL POWER, and W(z) is the cycle dependent function that
= accounts for power distribution transients encountered during normal operation. This function.is given in the Peaking Factor
~ Limit Report as per Specification 6.9.1.14. R21 -
t
- d. Measuring F (z) according to the following schedule:
- 1. Upon achieving equilibrium conditions after exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined,* or 9
- 2. At least once per 31 effective full power days, whichever occurs first.
*During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.
G E.ve 3/4 2 Ce R95 SEQUOYAH - UNIT 2 3/4 2-5 Amendment No. 21, 95 March 10, 1989
'y 'f '.
$If .A 5 q
e , s L ' t; POWER DISTRIBUTION LIMITSL !
/' "'
SURVEILLANCE REQUIREMENTS (Continued)-
, ;e, With' measurements indicating w . ..
g 1 maximum Fq (z) e over z K(z)i . ,
'has increased since the previous determination of F (z) either j ,. of the following actions shall be taken:
If . - 1. F (z)'shall be-increased by 2 percent over that specified in
)
;4.2.2.2.c,or~ .
R21
- 2. F (z)'shall-be measured at least'once per 7 effective full power days until-2 successive maps indicate that l
~ ~
maximum F (z) 'is not increasing. ,
, over z-K(z)
- f. With the relationships specified in 4.2.2.2.c above not being l satisfied:= .-!
e 1. Calculate the percent F (z) exceeds its limit by the
- l following expression: q F - =
p maximum Fn (z) x W(z) -1 x 100 for P > 0.5 R95 l L
'2.22?xK(z) )
,\('over[z ,P ,/ ,
e 2.32. .M. 9 f "%
- 'j maximum Fn (z) x W(z) I -1 ,x 100 for P < 0.5 R95 over z 0.22fxK(z) s 0. 5 J
- 2. - Either of the following actions shall be taken:
- a. Place the core in an equilibrium condition where the limit in 4.2.2.2.c is satisfied. Power level may i then be increased provided the AFD limits of Figure 3.2-1 are reduced 1% AFD for each percent R21 Fg (z) exceeded its limit, or
- b. Comply with the requirements of Specification 3.2.2 for F (z). exceeding its limit by the percent calculated above.0 R95 6 .. pay. ;/0 2 %.
SEQUOYAH - UNIT 2 3/4 2-6 Amendment No. 21, 95 March 10, 1989 l m
POWER DISTRIBUTION LIMITS g SURVEILLANCE REQUIREMENTS (Continued) 4 m g. The limits' .speci fied in 4.2.2.2.c, 4. 2.2. 2. e. and 4. 2. 2.2. f above - l are not applicable in the following core plane regions: J 1. . Lower core region 0 to 15 percent inclusive,
- c. 2. . Upper core region 85 to 100 percent inclusive. R21 j i
4.2.2.3 When Fq(z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured F (z) shall be obtained from a' ' 9
".. power distribution map and increased by 3 percent to account for manufacturing 1 tolerances further increased by 5 percent to account for measurement L - uncertainty. 'j 4
9 o , i l
- Dele +c I#The'l it shall 2.15 inste of 2.237 ntil an lysis in wit 10 CFR 50 6, using pl t operati conditi s and sho ng nformanc that a imit o 2.237 sa fies the r irements 10 CFR .46(b), h been com eted R95 nd submi ed to NRC.
SEQUOYAH - UNIT 2 3/4 2-6a Amendment No. 21, 95 March 10, 1989
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_ J{. /t_ V1 ! y I l 2 4 6 8 10 12 14 16 l-CORE HEIGHT ) -
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FIGURE 3.2 2. K(Z)- NORMALIZED F (Z) o AS A FUNCTION OF CORE HEIGHT \, 1 O SEQUOYAH - UNIT 2 3/4 2-7
~.___ . _ _ . . ..__ . _ _ . _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ . _ _ _ . _ . . _ _ . . . _ . , 77, . ,
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l e.es - 1 E S'etal Deaking Faster
- e. -
s.3: -
- m e.ee
- 9 900 S.900 6.900 1.900
- 10.000 '9.940
- 13.000 9 935 a.es - . a
. l e I e
g,, e e a e e.e a.e 4.0 e.e e.e es.o an.e Sees IIalght ' e e o t 9 9 i w *M-*-' ms mew _.__._____.-w- m ___-- m ___.____.___---__--_-_.___m ...._____.m_ _ . _ _ _
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T ABLE 3.4-1 H REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES , FUNCTION : VALVE NUMBER Accumulator Discharge L 63*S60 Accumulator Discharge 63-561 Accumulator Discharge ! 63-562 Accumulator Discharge 63-563 Accumulator Discharge 63-622 Accumulator Discharge
'63 623 Accumulator Discharge :
63-624
- Accumulator Discharge !
63 625 63 551 . Safety Injection (CoTo Leg) ! 63-553 Safety Injection (Cold Leg) , 63-557 Safety Injection (Cold Leg)
'63-555 Safety Injection (Cold Leg)
Residual Heat Removal (Cold Leg) , 63-632' Residual Heat Removal (Cold Leg) l 63 633 Residual Heat Removal (Cold Leg). 63-634 Residual Heat Removal (Cold Leg) . 63-635 Residual Heat Removal / Safety 63-641 Injection (Hot Leg) . Residual Heat Removal / Safety : 63-644 Injection (HotLeg) 63-558 Safety Injection (Hot Leg) , 63-559 Safety Injection (Hot Leg) . 63-543 Safety Injection (Hot Leg) ;
~
63-545 Safety Injection (Hot Leg) 63-547 Safety Injection (Hot Leg) 63-549 Safety Injection (Hot Leg) . ( Residual Heat Removal (Hot Leg) .. 63-640 Residual Heat Removal (Hot Leg) 63-643 _ ( 8 58 er Heaa In tion jection b 6O 87-560
-559 pper Head Upper H Injectio 87-5 Upper end Injec ' n
- 8 62 Up r Head In; tion 7-563 per Head section
- l. Upper He njection .
FCV (cha ng hende Uppe eadInje ion gyg. F 8* charging der) FCV-74-1* Residual Heat Removal FCV-74-2* Residual Heat Removal "These valves do not have to be leak tested following manual or automatic actuation or flow through the valve. 3/4 4-20 Amendment lio. 74
, SEQUOYAH - UNIT 2 September 21. 1988
1 ( O , e. j x 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMUtATORS i C010 LEG INJECTION ACCUMutATORS ' LIMITING CONDITION FOR OPERATION l 3.5.1.1 Each cold leg injection accumulato'r shall be OPERABLE with:
- a. The isolation valve open, t- ;
L b. A contained borated water volume of between and allons of
~
burated water.
' i s ': ed wi+6 . Between and ppm of boron, ; C sT ' r~emva I, ' and CMgf <*6 j g.p g g d. A nitrogen cover pressure of between M and psig. ;
been s AwW APPLICABILITY: MODES 1, 2 and 3.* Lin Ts change, es -20, ACT10N: l
- a. Withonecoldleginjectionaccumulatorinoperable,exceptasa t
( reiult of a closed isolation valve, restore the inoperable I accumulator to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the ' following 6 hours.
- b. With one cold leg injection accumulator inoperable due to the isolation valve being closed, either immediately open the isolation -
valve or be in HOT STANDBY within one hour and be in HOT SHUTDOWN within the next 12 hours. c.# With one pressure or water level channel inoperable per accumulator, return the inoperable channel to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours and in HOT R113 , SHUTDOWN within the following 6 hours, d.# With more than one channel (pressure or water level) inoperable per accumulator, immediately declare the affected accumulator (s) l inoperable.
" Pressurizer pressure above 1000 psig.
i #Cycle Actions e and doutage. 4 refueling are in effect until the restart of Unit 2 from the Unit 2 Ril3 SEQUOYAH - UNIT 2 3/4 5-1 Amendment No. 113 August !!, 1989
t
~
a
~. o- l This s ecification is delded. i Y ;
EMERGENCY CORE COOLING SYSTEMS L i UPPER HEAO, INJECTION ACCUMULATORS ' 1 LI ING CONDITION F0 PE:tATION / ! / . t
- 3. 5.1. 2 Each perheaainjectic accumulator.syste shall be OPERABLE ith:
I
; a. T isolation valve open, ;
- b. The water-fille accumulator contai ing between 1805 d 1851 cubic feet of borate water having a ce centration of be en 1900 and 2100 ppm of ron, and '
- c. The nitr en bearing accumu ter pressurized t between 1185 and ,
1285 p .g. APPtICABILITY MODES 1, 2 and ACT*0N: With the .:pper .ead injection ace .aulator system ine rable, except as a result a closed isolati valve (s), restor the Vpper head ( b. injet: tion cumulator system be in at east HOT STANDBY SHUT 00h within the folio ng 6 hours, Wit the upper head in'ection accumulat OPERAELE status thin the next 6 thin one hour or rs and in HOT ystem incperable ve to t isolation valve ( being closed, e r immediately op n the solation talve(s) r be in HOT STANC) within one hour d ce in - HOT SHUTDOWN wit n the next 12 hou 4. SU EILLANCE RE0VIREM WTS 1 1
/ .
4.5.1.2 Each perheadinjectio umulator syste' 11 be demonstr ed OPERABLE: /
- a. t least once per hours by:
- 1. Verifyin 4he contained bor ed water volume a nitrogen pressur in tne accumulato , and
- 2. Veri ying that each ac smulator isolatio valve is open.
"Pressuri resiure above 190 psig.
SEQUOYAH - UNIT 2 3/4 5-3 i
Deleto EMER@ftY CORE COOLING SYSTE/ EILLANCE REQUIRE ontinued) '
. b. least once per 31 dar and within 6 hours af each solution g
, volume increase of gr er than or equal to 1 ' of tank volume by verify ng the boro concentration of the 5 ution in the water fil d accumu ator,
- c. . At.least e per 18 months by:
- 1. erifying that each ac ulator isolation valv loses automa-tically when the wa level in the water-f ed accumulator is 92.0 + 2.6/-5.8 i es above the tank ye r working line when corrected for e mas's of cover gas.. RM
- 2. Verifyin hat the total dissolv nitrogen and air in
- water- lied accumulator is 1 s than 80 SCF per 1800 ubic j fe 'of water (equivalent 5 x 10 6 pounds nitro n per unds water).
- d. least once per 5 year y removing the membr installed between the water-filled and trogen bearing accumu ors and verifying that the removed rane bursts at a dif ential pressure of 40 + 10 psi.
i his pay in+entional13 Ieft- blanL . 9 4
n j
; EMERGENCY CORE COOLING SYSTEMS
'/
f
.I SURVEILLANCE RE'0VIREMENTS (Continued) k .. l
- h. By performing a flow balance test during shutdown following r
completion of modifications to the ECCS subsystem that alter the subsystem flow characteristics and verifying the following flow rates:
- 1. For safety injection pump lines with a single pump running:
- a. The sum of the injection line flow rates, excluding the highest flow rate is greater than or equal gpm, and R42
- b. The total pump flow rate is less than or eq o 675 gpm.
lR42
- 2. For centrifugal charging pump lines with a single pump running:
- a. The sum of the injection line flow rates, excluding the highest flow rate is greater than or equal gpm, and R42
- b. The total pump flow rate is less than or e ua to 555 gpm.
- 3. For all four cold leg injection lines with a single RHR pump running c flow rate greater than or equal to ,gpm.
31.51) ; December 1, 1986 SEQUOYAH - UNIT 2 3/4 5-8 Amendment No. 42 L -
4
+ -
e TABLE 3.6-1 (Continued) N BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING
- E SECONDARY CONTAIIMENT BYPASS LEAKAGE PATHS E
EE PENETRATION DESCRIPTION- RELEASE LOCATION X-92A,8 Hydrogen Analyzer Auxiliary Area b' X-93 Accumulator Sample Auxiliary Area
" Auxiliary Area X-94A,B,C Radiation Sample L. -
! X-95A,B,C Radiation Sample Auxiliary Area ~ T'7-s l X-96C Hot leg Sample Auxiliary Area X-98 ILRT Auxiliary Area
- X-99 Hydrogen Analyzer. Auxiliary Area L 79
- X-100 Hydrogen Analyzer Auxiliary Area f X-101 Postaccident Sampling, Containment Auxiliary Area X-103 Postaccident Sampling, Liquid Auxiliary Area Discharge to Containment R X-106 Postaccident Sampling, Air Auxiliary Arca
- R63 4 Discharge to Containment .
T X-108 - tHt!- main fenoncc. he W*'0" Auxiliary Area E X-109 -tMtf- Nin t ena"CC MC#"'#### Auxiliary Area X 110
'c_ ilia rj " n ;
X-114 Ice Condenser Auxiliary Area-X-115 Ice Condenser Auxiliary Area-X-116A Postaccident Sampling, Auxiliary Area Containment Air Sample
. ;e 4
ga< ?.
;a z
j 20 . i
, TABLE 3.6-2 (Continued)
E
.g CONTAllOENT ISOLATION VALVES i
) VALVE NUMER FUNCTION MXIIRM ISOLATION TIIE (Seconds) i g A. PHASE "A" ISOLATION (Cont.) lf Z
- 61. FCV-77-19 RCOT and PRT to V H
' u 10* .
- 62. FCV-77-20 Na to RCOT 10*
- 63. FCV-77-127 Floor Sump Pump Disch 10* .:
- 64. FCV-77-128 Floor Sump Pump Disch 10*
- 65. FCV-81-12 Primary Water Makeup 10*
l N@ 66. F 7 67 CV-87
~
UHI Test ine UHI T 10"f R62 j i j Line . , . FCV- UNI est Line 0*
- 9. FC 7-10 l'
l Test Li 10*
. 70. 11 I Test L' 10*
l k !
, 8. PM SE "B" ISOLATION >
U 1. FCV-32-81 Control Air Supply 10
- 2. FCV-32-103 Control' Air Supply 10
- 3. FCV-32-111 Control Air Supply 10
- 4. FCV-67-83 ERCW - LW Capt Cirs 60*
- 5. FCV-67-87 ERCW - LWR Capt Cirs 60* R29
- 6. FCV-67-88 ERCW - LWR Capt Cirs 60*
- 7. FCV-67-89** ERCW - LWR Capt Cirs 70*
f4
- 8. FCV-67-90** g73 ERCW - LWR Capt Cirs 70*
= 9. FCV-67-91 ERCW - LWR Capt Cirs 60*
k;g. 10. FCV-67-95 ERCW - LWR Capt Cirs 60* ga 11. FCV-67-% ERCW - LWR Capt Cirs 60*
,, 12. FCV-67-99 ERCW - LWR Capt Cirs 60* R29 .
P- 13. FCV-67-103 ERCW - LWR Capt Cirs 60*
- 14. FCV-67-104 ERCW - LWR Capt Cirs 60*
k;g 15. FCV-67-105** ERCW - LWR Capt Cirs 70* -
- 16. FCV-67-106** ERCW - LWR Capt Cirs
$ 17. FCV-67-107
- 18. FCV-67-111 ERCW - LWR Capt Cirs ERCW - LWR Capt Cirs 70*
60* I R73 60*
$ 19. FCV-67-112
- 20. FCV-67-130 ERCW - LWR Capt Cirs 60* '
y w ERCW - Up Capt Cirs. 60e R29
- 21. FCV-67-131.
cn EACW - up Capt Cirs - 60* - - . -
- 60* ,
- 22. FCV-67-133
'ERCW ' $ Capt Cirs' '
, s
. ...,y ..v _ . , , , -. ,,
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l i 3/4'.2 POWER DISTRIBUTION LIMITS
~ !
BASES ~ i
. The specifications of this section provide assurance of fuel integrity l during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) i events by: (a) maintaining the calculated DNBR in the core at or above design +
during normal operation and in short term transients, and (b) limiting the ,
. fission gas release, fuel pellet temperature and cladding mechanical properties '
to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial -l conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded. The definitions of certain hot channel and peaking factors as used in , these specifications are as follows: R 21 *
.F (Z) Heat Flux Hot Channel Factor, is defined as the maximum local ,
O heat flux on the surface of a fuel rod at core elevation Z divided ' by the average fuel rod heat flux, allowing for manufacturing ; tolerances on fuel pellets and rods. N F Nuclear Enthalpy Rise Hot Channel Factor is defined as the ratio of the 1Negral of linear power along the rod with the highest integrated power to (. the average rod power. , 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) The li AXIAL FLUX DIFFERENCE assure that the qF (2) upper bound envelope of . times the normalized axial peaking factor is not exceeded L during either normal operation or in the. eve'nt of xenon redistribution follow-ing power changes. 1 ! Provisions for monitoring the AFD on an automatic basis are derived from l the plant process computer through the AFD Monitor Alarm. The computer deter-mines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the allowed AI-Power operating space and the THERMAL POWER is greater than 50 percent of RATED THERMAL POWER. l 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, RCS FLOWRATE AND , NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR
, The limits on heat flux hot channel factor, RCS flowrate, and nuclear enthalpy rise hot channel factor ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.
SEQUOYAH - UNIT 2 B 3/4 2-1 Amendment No. 21
~SEP 2 01983
in the cpecificab:n +kr occumulator' volum and not (The. coeer Simihpr3ssure. - cre en3Iysis' limits Gnd O* *** in"f3 *n5ff"m!7erf \i
. ne etwr precsura limha wert d2iermiz:d by Wes&ghusc. 6 b uncerninhtend be ' Cts ps c'ti.6 psia . Sines the. instrumes+ ee d-evis in th'. confre/
r** m ars in psip the T5 values have been conve.rted to psi 3 and rounded l t to ihe nenrest whele numbers. ne. aciuel nifeopen corer pres.sfre, safefy limits ( 3/4.5 EMERGENCY CORE COOLING SYSTEMS
' in S4N's dess'4n documen/s are 6 00. 3 psi 3 an"d 662.s pso .
L BASES' , 3/4.5.1 ACCUMULATORS j The OPERABILITY of each cold leg injection Ia [ u d A ead 4 daefihri OC./C[C, accumulator ensures that a sufficient vq1ume of borated water will be L immediately forced into the reactor core in the event the RCS pressure falls : below the pressure of the accumulators. For the cold leg i } ion accumulators this condition occurs in the event of a largeA u ure.f u on t for y th thy arge g d sma p breaky uue its y , bC : i l The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.[hdci INSERT 8] ene accumuiswr power operated isolation valves are considered to be
" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive ,
conditions are not met. In addition, as these accumulator isolation valves L . f ail to meet single f ailure criteria, removal of power to the velves is required. , The limits for operation with an accumulator inoperable for any reason L except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator l which may result in unacceptable peak cladding temperatures. If a closed l isolation valve cannot be immediately opened, the full capability of one accumulator is not available 'and prompt action is required to place the reactor in a mode where this capability is not required. 3/4'5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA . l assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each-ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period. , l With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is ! acceptable without single failure consideration on the basis of the stable l reactivity condition of the reactor and the limited core cooling requirements. r L. l SEQU9YAH - UNIT 2 B 3/4 5-1
i e . IIVSER T B t The rninim uwt boron concenirulion ensures Ura.6 the i reac+or core remain ' will sobcnWcaI duriny the acc urn ula tor inj ec+ ion period of a small break
~
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^
m ; , l l 1 ENCLOSURE 2 l l PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 ) i DOCKET NOS. 50-327 AND 50-328 ; (TVA-SQN-TS-89-25) p DESCRIPTION AND JUSTIFICATION FOR [ .. UPPER READ INJECTION REMOVAL j
. i 1
f, I 4 6 4 i e
}
L l l l l _ _ . _ . .- ._ . _ _ .
e ENCLOSURE 2 Description of Chante Tennessee Valley Authority proposes to modify the Sequoyah Nuclear Plant (SQN) Units 1 and 2 technical specifications (TSs) to reflect the removal of the upper head injection (UHI) system. TS 3/4 5.1.2 has been deleted, and Tables 3.4-1, 3.6-1, and 3.6-2 for reactor coolant system (RCS) + pressure isolation valves, penetrations, and containment isolation valves have been revised. The peaking factor limit has been changed in Limiting Condition for Operation (LCO) 3.2.2 and Surveillance Requirement , (SR).4.2.2.2. As a result of the peaking factor revision, Figure 3.2-2 : has also been revised. LCO 3.5.1.1 for the cold les injection L accumulators reflects new required values for volume of water and nitrogen ! cover pressure. Changes have also been made to minimum flow rate values in SR 4.5.2.h. Reason for Channe The proposed TS changes are a result of the removal of the UHI system and the new analyses performed to support their removal. Because of the complex and rigid requirements placed on its performance, [ the UHI system has introduced concerns regarding appropriate water volume , delivery, nitrogen injection into the RCS, fluid mixing behavior, and an ' increased number of plant mode changes involved with UHI problems and their resolution. Significant maintenance is required for the UHI system, particularly for the UHI main isolation valves. ALARA (as low as reasonably achievable) concerns are involved with system maintenance, as well as with the increased time for reactor vessel head removal because of URI connections and piping. The UBI system has also been the subject of regulatory concerns at SQN, including the integrated design inspection and two licensee event reports. TVA committed to NRC in a November 3, 1988, letter'to remove the UHI system before restart from the Unit 1 and Unit 2 Cycle 4 refueling outages.
~ Justification for Channe 4 l
For UHI removal, Westinghouse Electric Corporation has reanalysed depressurization of the main steam system, main steam line rupture, small : break loss of coo' ant accident (LOCA), and large break LOCA for SQN. In addition to removal of the UHI system, TVA plans to implement modifications during each unit's Cycle 4 refueling outage that will implement Eagle 21 digital protection, resistance temperature detector bypass elimination, boron injection tank deactivation, new steamline break protection, Vantage 5 hybrid fuel upgrade, and reactor trip on steam flow / feed flow mismatch. These modifications have been evaluated for impact on the analyses performed to support removal of the UHI system. l I
.. -_ -. -- - - - .- ~
4
.g.
With'a change in the cold leg injection accumulator volume of water and an increase in the accumulator nitrogen cover pressure, the new analyses support the removal of the UHI system. The cold leg accumulator nitrogen cover pressure limits were determined by Westinghouse Electric Corporation ; to be 615 pounds per square inch absolute (psia) and 697.5 psia. Since the instrument read-outs in the main control room are in pounds per square inch gage (psig) TVA has converted the TS values to-psig and rounded to ' the nearest whole numbers. Inst'rument inaccuracies have been considered
- in the alarm setpoints but are not reflected in the TS values.
The removal of the UEI system results in the deletion of the URI RCS
. pressure isolation valves and the UHI test line containment isolation ;
valves. Penetration X-110 is also being deleted; it is being capped with ' a socket welded fitting on the outboard side of containment. Penetrations .
-X-108 and X-109 will be_ converted to maintenance penetrations and will be fitted with testable double 0-ring blind flanges on the inboard side of containment.
As a result of improved modeling techniques, the analyses performed to support UNI removal resulted in changes to the peaking factor limit and
-minimum emergency core cooling system flow rates. The peaking fector limit is increased from 2.237 to 2.32, and the footnote that currently limits the peaking factor to 2.15 is deleted. The minimum value for the l sum of the safety injection pump line flow rates decreases to 443 gallons -
per minute (spm), and the minimum value for the sum of the centrifugal > charging pump line flow rates decreases to 309 gpm. The minimum flow rate , for all-four cold leg injection lines decreases to 3931 spm. . To summarize, Westinghouse reviewed all Final Safety Analysis Report (FSAR) Chapter 15 analyses and determined that depressurisation of the main steam system, main steamline rupture, small break LOCA, and large break LOCA should be reevaluated for UH1 removal. With compensating changes to the cold leg accumulators, the new Chapter 15 analyses support UHI removal. Environmental Impact Evaluation The proposed change request does not involve an unreviewed environmental question because operation of SQN Units 1 and 2 in accordance with this change would nott l
- 1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the Staff's testimony to the Atomic Safety and Licensing Board, supplements to the FES, environmental impact appraisals, or in any decisions of the Atomic Safety and Licensing Board.
- 2. Result in a significant change in effluents or power levels.
- 3. Result in matters not previously reviewed in the licensing basis for SQN that may have a significant environmental impact.
g q,7 i l' i r ! r- <
-ENCLOSURE 3 1
- . , , - . . . :.w , ---
l PROPOSED TECHNICAL SPECIFICATION CHANGE j SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2' DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-89-25) DETERMINATION OF NO SIGNIFICANT RAZARDS CONSIDERATIONS . b f 1 I h 1 L e y , - . . . . - _
7- ; i ENCLOSURE 3 j i Significant Hazards Evaluation I TVA has evaluated the proposed TS change and has determined that it does ; notLrepresent a significant hazards consideration based on criteria t established in 10 CFR 50.92(c). Operation of SQN in accordance with the ! proposed amendment will nott ; (1) Involve a significant increase in the probability or consequences of an accident previously evaluated. Westinghouse has evaluated all FSAR Chapter 15 analyses to determine which analyses are affected by UHI removal. Main steamline rupture, small break LOCA, large break LOCA, and depressurization of the main : steam system were identified. A reanalysis of these events confirms that with compensating changes to cold leg accumulators, the design : bases continue to be met without URI. Therefore, the removal of the UHI system does not involve a significant increase in the probability ; or consequences of an accident previously analysed. (2) Create the possibility of a new or different kind of accident from ! any previously analyzed. UNI was designed and installed at SQN as a mitigation system and was never assumed to initiate an event. Therefore, UHI removal should not initiate a new or different type of accident. The modification j will leave four penetrations of the reactor head that will be j_ permanently sealed and tested. Any future leakage from the i l penetrations would be bounded by SQN's small break LOCA or large break LOCA analyses. No new or different kind of accident from any previously analyzed is anticipated. l (3) Involve a significant reduction in a margin of safety. , Westinghouse has evaluated all FSAR Chapter 15 analyses to determine which analyses are affected by UHI removal. Main steamline rupture, l .- small break LOCA, large break LOCA, and depressurization of the main i steam system were identified. A reanalysis of these events confirms
- l. that, with compensating changes to the cold leg accumulators, the i design bases continue to be met without UHI. Since the design bases contain the required margins of safety, the removal of the UHI system l does not involve a significant reduction in safety margins.
l l l I
g,7 - - . - s 3
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. ENCLOSURE 4-Final Safety Analysis Report
. Chapter 15 Analyses Expected Changes d
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-_-.__m_ _ - . _____ _ _ __ _____m___ . . . . , . , , - , , , , _ _ . ,. , , p. . - v
t SQN-6 . i. i ,
.I
. . .ABLE OF CONTENTS (Continued)
' )I'
,J .,
Section Title M > 15.3-10 ;
.15.3.5 WASTE GAS DECAY TANK RUPTURE' 15.3 10 15.3.5.1' Identification of Causes and Accident Description
' ~ Analysis of Ef'fects and Consequences 15.3-10 1 L 15.3.5.2 15.3-10 l 15.3.6 SINGLE R00 CLUSTER CONTROL ASSEMBLY WITH0RAHAL AT i
FULL POWER i IdenDfication of Causes and Accident Description 15.,3-10 j 15.3.6.1 15.3-11 15.3.6.2 Analysis of Effects and Consequences Conclusions 15.3-12 L 15.3.6.3 15.3-12 l 15.
3.7 REFERENCES
15.4,1
- 15.4 CONDITION IV - LIMITING FAULTS 15.4-1 15.4.1 MAJOR REACTOR COOLANT SYSTEM
- PIPE RUPTURES (LOSS OF COOLANT ACCIDENT) t Thermal Analysis F- 15.4-2 15.4.1.1 .
15.4.1.1.1 Westinghouse Performance criteria for Emergency L5.4 2 y Core Cooling System
-Method of Thermal Analysis 15.4-3 15.4.1.1 1 15.4-3 15.4.1.1.3 Containment Analysis ,
Results _ of Large Break--Spectrum . ~ 15.4.-3
.. 15.4.1.1.4
}}q.}.).} , E((ecp g ontainment Purg{ng ,, ,__u_
gym gW ' , ,
'('
, . ..... . . . . . . . . . . _ .........--..~......
. Thermal Analysis 15s4-8
)- 5._ . . 7 Conclusion -
15.4-9 15.4.1.Z Hydrogen Production and Attumulation 15.4.1.2.1 6 #'IL Method of'. Analysis 15.4-9 15.4.1.2.2 7~ Assumptions" - 15.4-9 *
't Core Solution Radiolysis - 15.4-11 ;
15.4.1.2.3 ' 15.4-13 15.4.1.2.4 E"'# Sump S61ution Radtplysis PLimary Coolant Hydrogen 15.4-l.4 15.4.1.2.5 15.4-14 1-15.4.1.2.6 .
- Results of -Analysis -
MAJOR SECON'0ARY SYSTEM PIPE RUPTURE 15.4-14
. .15.4.2 15.4-14 15.4.'2.6 Rup%re of a Main Steam Line Identificat4on of_fcauses and Accident Description 15.4-14 15.4.2.1.1 15,4-16 15'4.2.1.2 Analysis of Effects and Consequences .
M4jor Rupture of a Main Feedwater Pipe 15.4-28 15.4.2.2 15'.4-28 15.4.2.2.1 3dentification of Causes and Accident Description Analysis of Effects and Consequences 15.4-29 7. s 15.4.2.2.2 , 15.4-32
- 15.4.2.2.3 ., Conclusion .
l
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15.'4.32 l 15.4.3 STEAM GENERATOR TOBE RUPTURE 15.4-32 15.4.3.1 Identification of Causes and Accident Description Analysis of Effects and Consequences 1E.4-33 15.4.3.2 :
- 15.4-36 15.4.3.3 -
C~onclusions
'15.4-37 e 15.4.4 SINGLE REACTOR COOLANT PUMP LOCKED ROTOR ~
15.4.4.1 Identification of Causes 'and Acc1 dent Description ~ ,15.4-37 ,. e
~
15-4 1 t)113F /COC4 l I { . . _ _ _ _f _.;_-_ _____ -_ s'_ [ _
. a
t t
~
. SQN-1
. r.
L.IST OF TABLES l(~
.(-
~
Number- Title 15.1.2-1 Nuclear Steam Supply System' Power Ratings 15.1.2-2 ~ Summary of Inittil Conditions and Computer Codes , 15.1.3-1 Trip Points and Time Delays to Trip Assumed in Accident Analyses .. ; 15.1.4-1 Determination of Maximum Overpower Trip Point-Power Range Neutron Flux Channel - Based on Nominal Setpoint Considering Inherent Instrumentation Errors - 15.1.7-1 Core and Gap Activities Based on Full-Power Operation for 1000 Days - Full Power: 3565 MWt , 15.1.7-2 Core
- Temperature Distribution '
'~-
15.2-1 Time Sequence of Events for Condition II. Events _, , 15.2'.7-1 Minimum Calculated DNBR for Cases of Rod Cluster Control Assembly MisalignmentandDroppedRodClusterControlAssembfy 15.2.4-1 , Sequence of. Events ,
~
15.2.4-2 SequenceofEveNs - a. ! 15.2.7-1 Initial Conditions for a Complete loss of Load from 52% Power
- 15.2.7-2 ~ .r. Time Sequence of. Events for a Turbine Trip.at 52% Power with '
Pressurt'rer Pressure Control
- n. '
. 15.2.7-3. - Time Sequence of Events for a Turbine Trip at 52% Pow'er without l ,, Pressuri-ze'r Pressure Control . 15.2.9-1 NaturalCirculatiofFlow
- 15.3.1-1 4LoopSmallBreakTimeSequenceofEvents
~
~
4 Loop.Small Break ; 15.3.1-2 15.3.4-1 fime Sequence of Events for Condition III Events . 15.4.1-1 , Large Break kt/MdhDEtert , Largt: Brea'It- 10% SGTP A' N L 15.4.1-14 .5 PEtraum er, h.un. ar
~ ..
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- l 15 6 0113F/C0C4
'l ;
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~
L157 0F TARLES (Continued) . g {' g c ' O :-: 1. - .. :..: : ::.; C . 15.4.1-3 BackpressureTranstentUsedinknalysts 15.4.1-4 Containment Data Redired for ECCS Evaluation - Ice Condenser - Contalament 15.4.1-5 Major Characteristies of structural Heat links Ins 1de seguoyah Nuclear Plant Containment - 15.4.1-6 Mass and Energy Release Rates. C.. O. !-:-?. @ , Q j 15.4.1-7 ' .. Large treak Time 'SeeveAce__9f_EyerttLk f ' -t. "'.d...gb
- .
- .:.: :. -- . ;. n.e. := &!". != n:= n e :....:.. :,.cra: -n:EM I
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1 9 .!
- M 7;; " :2 . 'j = 0 ;r x 5" S: t: ": '--t "' '"a ,
~ ~
W.'" "h:t " : :tx; '?xd ' '_M'. ^^:!;;;': # l 15.4.1 10 Nodal + Representation of Core Transients , .
~
L '* !' 15.4.1-11 Variation'UN!NaterVolume(Deliveredi i
'.'15.4.1-12 Time Sequence of Ev,ents for Con'dttion iV Events -
c-15.4.1-13 Post Accident containment Temperature Transient Used in the ' Calculation of Aluminum Corrosion
~
~ . .
15.4.1-14 Parameters Used to De'termine Hydrogen Generation l
~ '
l 15.4.1-15 , core F'thton Produci Energy Af ter 830 Full Power Days
~ 15.4.1-16 F1ss1ovi Product Decay Deposition in 5'mp u Solut1on -
. . 4 15.4.2-1 Core Parameters used 1trSteam treak DNB Analysis 15.4.4 Summary Df Results for Locked _ Rotor Transtants , _-
15.4.6-1 . Parameters Used in the Analysts of the Rod cluster control ' t
.- -Assembly Ejection Accident l ~
15.5.1-1 .: Parameters Used in Loss.of A.C. Power Analyses _.
.,, 15.5.2 1 Parameters Used in Maste Gas Decay Tank Rupture Analyses l . . .
15.5.2-2 Naste GasD ecay Tank Inventory (one untt) .' W
~
15-7 - 0113F/COC4 d. L 095 . .
u, S~ : u - ; j
- SQN-6 i .,
L( LIST OF FIGURES l
l Number' Title :
L 15.1.3-1 illustrationofOvertemperatureandOverpowerATProtection 15.1.5-1 ~ Control Rod Pott'tlon Versus Time on Reactor Trip j
' 15.1.5-2 NormalizedRCCAReactivitybrthversusRodInsertion .
l 15.1.5-3 Normalized RCCA Bank Reactivity Worth versus Time After Trip , j 15.1.6-l' Doppler Power Coefficient Used in Accident Analysis - Residual Decay Heat 15.1.8-1 s- - 15.1.9-l' Tuel Rod Cross Section , 15.1.9-2 Nuclear Power Following-Trip for Input to the BLKOUT Code
',.,'^' ";;0 ":M hje;t L. 0,ete; ;;;.eeh% DE( EPc" 15.2.1-1 -Uncontrolled Rod Withdrawal From a Subcritical Copd,ltion Neutron Flux,versus Time _
~
g s 15.2.1-2' Uncont' rolled Rod Withdrawal From a Subtritical Condition Thermal
.( l( Flux versus Time -
c.
- . 15.2.1-3' Un. controlled Rod Withdrawal From a Subscritical Condition, * * '
l Temperature versus -Time, Reactivity Insertion Rate 75 x 10* l' - AK/sec- ._ l 15.2.2-1 Typ_lcal Transient Response for Uncontrolled Rod Withdrawal from j
- Full Power Terminated by High Neutron Flux Trip
[
- '15,t.2-2 Typtu l Transient Response for Uncontrolled Rod Withdrawal From
- FullPowerTerminejedbyHighNeutronFluxTrip 15.2. 2-3 Typ_lca1' Transient Response for Uncontrolled Rod Withdrawal from Full Power Terminated by.0vertemperature AT Trip _'
15.2.2-4 Typicai Transient Response for Uncontrolled Rod Withdrawal From
- Iull Power Terminated by Overtemptrature AT Trip ~.
~ '
15.2.2-5 Ef.fect of Reactivity Insertion Rate on Minimum DNBR fon a Rod
- Wilhdrawal, Accident from 100% Power .
15.2.2-6 Effect of Reactivity Insertion Rate on Minimum DNBR for a Rod ,, i Withdrawal Accident from 60% Power .
. s .
l
- 15-9 0113F/COC4 i 4
$ - L _2E H~i W ' A -~* G b~!Y
.s !
SON-6 f - - LIST OF FIGURES (Continued) N,,ymjter, ., Title 15.3.3-5 Loading a Region 2 Assembly Into a Region 1 Position Near Core Periphery .,
'~X11 Loops Operattig. All Loops Coasting Down, Flow Coastdown l 15.3.4-1 ~
. versus Time .
15.3.4-2 All Loops Operating, All Loops Consting Down, Flux Transients 15.3.4-3 . All Loops Operating..All Loops Coasting Down, DNBR versus Time ,
~
15.3.414 , All But one Looo Operatlng. All Loops Coasting Down, flow Coastdown versus Time - 15.3.4-5 AIT But One Loop Operating, 411 Loops Coasting Down, Flux Transients s_, DNBR versus' Time All But One Loop Operating All Loops Coasting ? 15.3.4-6
.. Down ,
t Compartment Pressure 15.4.1-1 ,
* +
'15.4.1-2 . Typical SATAN Model for a PWR (53 Elements) UHI'
,l 15.4.1-3 -
Fluid Quality - CECLG (C. . 8),fimp Mlx- '
- Burst Elev 645 Feet, Peak Elev 7.25 feet 15.4.1-4 Fluid Quallky - DECLG'(C. 0.8), Per Mix '
~ Burst Elev 6.0 feet, Peak Elev 6725 feet .
0.6), Imp Mix - ), 15.4.1-5 Flut4-Quality - DECLG (C. *
- Surst Elev 4.2'5 feet, Peak 'Elev 7'25 feet
- ~ -
. ReNsE Fluid I)Uallty - DECLG (C. 0.4), Imp Mix 15.4.T-6 mrd Burst Elev 5.75 Fee~ R Peak Elev 7.5 feet fMset7 C. -
15.4.1-7 Mass Velocity - DECLG (C. 0.8), Imp Mix , l . Burst Elev 6.25 feet, Peak-Elev 7.25 Feet _
. 4
{ ' 0.8), Ptr Mix
~
15.4.1-8 Masr Velocity - DECLG (C. . l. Burst E]ev'6.0 feet, Peak Elev 6.25 feet 0.6). Imp Mix 15.4.1-9 ' Mass. Velocity - DECLG (C. Burst Elev 6.25 feet, Peak Elev 7.25' Feet 15.4.1-1Q Mass Velocity - DECLG (C. - 0.4). Imp Mi'x . Burst Elev 5.75 feet, Peak Elev 7.5 feet - 4 (* e.
~
0113F/COC4
' 15-1E ..
6 l l b' .dmu <s* n .
- T. %
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*** INSERT C ***
1 C . l
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15.4.1-2 RCS Pressure - DEC14 , C D=0.6 j
- 9 .;
15.4.1-3 Core Flowrate - DECLG , C D=0.6 j l 15.4.1-4 . Cold Ler Accumulator Flowrate - DECIA , CD=0.6
.15.4.1-5 Core Pressure Drop - DECIA , CD =0.6
;15.4.1-6. Break Mass Flowrat,e - DECIA. , CD=0.6 I
15.4.1-7 Break Energy Flowrate -DECLG , Cp=0.6 . . l 15.4.1-8 Normalized Core Power - DEC'14 , Cp=0.6 . l
' " i 15.4.1-9 Core and Downconer Liqu'id Levels - DECLG ., CD=A 6 M' .
15.4.1-10'. Core. Inlet Fluid velocity - DECI.G ,C (as input to the' thermal analysis cod!=) 0.6 . Accumulat.or and Pumped Safety. Injection Flowrates 5
~
,15.4.1-11
,. DECLG ,'Cp=0.6 . .
15.4.1-12 Fluid Mass Flux - DECLG ,"ND=0.6 ,
. . - - \
- l. 15.4.1-13 Rod Heat Transfer coefficient J DECLG , Cp=0.6 i Clad Peak Temperature - DECLG , CD=0.6 15.4.1 l
- u
' ~
15.4.1-15 .CladTemperatureat-theBurst} Location-DECLG,C=0.6 p 15.4.1-14 Upper Compartment-S.tructural Heat Removal Rate - DECLG , C D=0.6- 7 15.4.1-17. Lower, Compartment structural Heat' Removal Rate - DECLG , CD=0.6 . 15.4.1-18 ' Compartment, Temperature - DECLG , Cp=0.6 , 15.4.1-19 , Heat Removal by' Sump - DECIA ,pp=0.6 _ 15.4.1-20 N'est Ramoval by IC Drain - DECLG , CD =0.6 t O
~
r- - 057 i ..
1 O , SQN-6 L LIST OF FIGURES (Continued)
'- '( - er Title
$.675
~
- 15. 1-11 Heat Transfer Coefficient - DECLG (C. 0.8) Imp Mix Burst Elev 6.25 fee,t, Peak Clev 7.25 feet
-- v.
), Per Mix 15.4.1- Heat-Transfer Coefficient - DECLG (C. 0 Burst Elev 6.0 Feet, Peak Elev 6.25 fee .. !
e 15.4.1-13 eat Transfer Coefficient - DECLG (C. - 0.6), Imp Mix rst Elev 6.25 Feet, Peak Elev 7.25 eet
. l
. 15.4.1-14 Heat Transfer Coeffletent, DECLG . - 0.4) Imp Mix !
r 3 ; 15.4.1-15 RCS Pre sure, DECLG (C., = 0.8), mp Mix , ! i 15.4.1 RCS Pressu , DECLG (C. . 0, ,
,Per Mix v- .
.6), Imp Mix 15.4.1-17 RCS Pressure, ECLG (C. = ,, y 15.4.12t8 RCS. Pressure D LG (C. 0.4), Imp Mix !
l 15.4.1-19 Core Flow Rate, DE G C.'_= 0.8) Imp Mix Lower Half 'of Core,,
- 15.4.1-20 ~ Core Flow' Rate, DEC G . = Q.8) Per. Mix Lower Half of Core
~
15.4.1-21 Core Flow Rate, CLG (C. G.f ) , Imp Mix' Lower Half of Core -
- 4) Imp'M1x Lower Half of Core 15.4.1-22 Core' Flow Rate DECLG -(C. .
0.8 Imp Mix Upper Half of Core 15.4.1-23 Core Flow Ra e, DECLG (C. 15.4._1-24 , Cordilow atei DECLG (C. 0.8) Pe Mix Upper Half of Core .- L. ' j 15.4lE25 Core F1 Rate, DECLG=(C. 0.6) Imp x Upper Half of Core >
= .
15.4.1-26 Core low Rate, DECLG (C. 0.4) Imp Mi'x per Half of Core L 15.4.1~27 Vo d . fraction e DEC'LG (C. =_0.8) Imp Nix Lowe Half of Core -
- 7. -
15.4.1-28 old. Fraction, DECLG (C. 0.8), Per Mix Lower if of Core < ! 15.4.1-29 Vold Fraction, DECLG (C. 0.6), Imp Mix Lower Hal of Core l 15.4.1- . Void iraction,.DECLG (C. - 0.4) Imp M1x Lower Half of or'e
'15.4. -31 void Fraction, DECLG-(C. 0.8) Imp Mix Upper Half;of Co , ,
15 .1-32 Vold Fraction, DECLG (C. = O.8) Per Mix Upper Half of Core l 5.4.1-33 ' V,old Fraction, DECLG (C. 0.6) I'mp Mix Upper Half of Core, i l g 4 . 15-15^ 0113F/COC4 '; F
- h. -
l SON-6
._ l LIST OF FIGURES (Continued) t
' s ;
Title j(( vmber
- 15. 1-34 ,Nold Fraction, DECLG (C. 0.4) Imp Mix Upper Half of Core i
15.4.1 35 Peak Cladding Temperature, DECLG (C. - 0.8) Imp ix ,
?
Surst Elev 6.25 feet. Peak'Elev 7.25 Feet 0.8) er Mtx 15.4.1-36 Peak Cladding Temperat'ure, DECLG-(C. urst Ele,v 6.0 Feet, Peak Elev 6.25 Feet . 4 t Cladding Temperature, DECLG (C. 0 ) Imp Mix ! 15.4.1-36a Pe
, Bur Elev 6.25 Feet, Peak Elev 7.25 Fe 15.4.1-37 Peak C dding Tamperature, DECLG (C. 0.4) Imp Mix ,
Burst El V 5.75 Feet (Peak Elev 7.5 eet 15.4.1-38 Fluid Tempe ature, DECLG (C. - 0. Imp Mix Burst Elev 6 5 Feet, Peak Elev .25 Feet - hh 15.4.1-39 Fluid Temperit e,DECLG[(,. .8) Per Mix " s Burst Elev 6.0 F et, Peak Ele 6.25 Feet +
= 0.6) Imp Mix 15.4.1-40. . Fluid Temperature, ECLG ( .
Burst Elev 6.25 Feet, Pea lev 7.25 Feet 15.4.1-41 Fluid Temperature,. DECL (C. - 0.1)~ Imp M - Elev 7.5 Feet',lx Durst Elev 5.75 Feet, e - .j . ._
. . s..
15.4.1-42 'Reflood Trans' tent, CLG (C = 0.8) Imp Mix , 15.4.1-43 'Reflood Transient DECLG (C. . .8) Per Mix _ 15.4.1-44 Refloed Transie t4, DECLG (C. . O. ) Imp Mix .
. 1 15.4.1345
~
,Reflood Trari' ent, DECLG (C. 0.4). mp Mix l 15.4.1:.46 Reflood Ra e, DECLG te. 0.8) Imp Mix l
.Reflood ate, OECLG (C. 0.8) Per Mix l 15.4.1-47 15.4.1-48 Re f'1 d Rate, DECLG '(C. 0.6) Imp Mix
^
15.4.1-49 Ref ood Rate, DECLG (C. 0.4) Imp Mix . 15.4.1-50; cumulator Flows,-DECLG (C. - 0.8) Imp Mix _ , I
.15.4.1-51 Accumulator. Flows, DECLG (C. = 0.8) Per Mtx
\ I 15.4.1-5 Accumulator Flows, DECLG (C,, = 0.6) Imp Mix.' [ .
'15.4.1 53 Accumulator FI'o'ws, DECLG (C. 0.4) Imp Mix .
'~
15-16g ~ 0113F/COC4 % 0 0
# ap 4 g
e ,. : y ' 4 ' .
- SON-6 c.
7 , , LIST OF FIGURES (Continued)
^
r Title .
- 15. 1-54 SI + Accumulator flow, DECLG'(C. = 0.8) I p Mix
- 1. .
15.4. 55 SI + Accumulator" Flow, DECLG (C.. 0.8 Per Mix (
'15.4.1-5 SI + Aqumulator Flow DECLG (C. 0 6) Imp Mix -;
15.4.1-57 SI + Accumulator _ Flow, DECLG (C. 0.4) Imp Mtx . 15.4.1-58' leted by Amendment 1-15.4.1-59 Del d by Ambdment 1 (({ [ 15.4.1-60 061ete by Amendment 1 ,
^
15.4.1-61 Delete'd y , Amendment 1
~
~
15.4 1-62 Lower Compar ment Str ctural Heat Removal Rate 15.4.1-63 Ice. Condenser al Flow versus Time OECLG (C. 0.8). Imp Mix
~~
~15.4.1-64 Compartment Temp ture '
v( 15.4.1-65 Heat Removal. b -Sump - a. 15.4.1-66 ~ Heat Removai by IC Ora ,
- 15.4.1-67 ~ JUpper C artmentiStructub.1 Heat- Removal Rate _
15.4.1-68 Flukd. ality, DECLG (C . O. ) Imp Mix, 10% SGTP . .
~
- Burst lev,6.5 feet,. Peak Ele 7:25 feet
^
15.4ft-69 Flu o allty,' DECLG.(C.~~ - 0.6) p Mix, 10%.SGTP l. l - Bu st Elev 6.75 Fefr, Peak Elev 7. Feet
~
15.4 J -70 ass Velocity, DECLG (C. - 0.8) Imp M , 10% SGTP
- Burst Elev 6;5 Feet, Peak-Elev 7.25 Fee
- 7. ,i 0.6) -Imp Mix,1
- SGTP 'l 15.4.1-71 M4ss Velocity, DECLG (C. ,
L ' Burst Elev 6.75 Feet, Peak Elev 7.25 Feet - ) ~
- ~
15.4. -72 Heal Transfer Coefficient, DECLG (C. 0.8) Imp x, 101 SGTP Burst Elev 6.5 feet, Peak Elev 7.25 Feet l .4.1-73 Heat Transfer Coefficient, DECLG (C. 0.6) Imp Mtx, % SGTP .. Burst Elev 6.75 Feet, . Peak Elev 7.25 feet . . 15.4.1-74 RCS Pressure,"OECLG (C. - 0.8) 10% SGTP - T' - - 15-17 0113F/COC4 i
~
D .: a . . ......:.
mp i# SQN-6
- - LIST OF FIGURES (Continued) i 4(' )
k^ Jhtskr Title .s
.. { >
15.4.1-75 RCSPressu/,DECLG(C = 0.6) 1 SGTP-15.4.1-7 Flow Ra . DECLG (C . 0.8) Im Mix 10% SGTP Lower Half of ore 15.4.1777 Flow ate, DECLG . . 0.6) mp Mix 10% SG . Lower Half o Core
-15 d-78 F1 Rate, DEC (C. . O. 10% SGTP Up r Half of Core Iow Rate. O CLG (C. ,= .6) 10% SGTP pper Half of Cor-e 15.4.1-79 .
15.4.1-80 on. :BECLG ,(C. 0.8) 10 SGTP Lower Half of Core 15.4.1-8 /.VoldFrac Voi,d Fr ction, DE G (C. 0.6) 0% SGTP Lower Ha f of Core . 15.4.1'82 Void raction, ECLG (C. ='O. ) 10% SGTP Upper Half of Core F~ 15.4 1-83 Vr d Fractio , DECLG (C. . .6) 10% SGTP Up r Half of Co y 1 .4.1d4 eak Cladd ng Temperatur . DECLG (C. = 0. ) Imp Mix, 10 SGTP Burst El 6.5 feet, P k Elev 7.25 fee , f ., dding Temper ue DECLG (C. 0.6) Imp Mt , 10% SGEP 15.4.1-85 . Peak Buts lev C.75 Fee , Peak.Elev 7.2 feet . d' Flu ' Temperatur DECLG (C. '=' 0 )' Imp Mix, 1 SGTP 15.4? 86 B st Elev 6.5 F et, Peak Elev. 25 feet , 15 4.1-87 luid Tempera ure. 0ECLG (C. 0.6) Imp M ,-10% SGTP " Burst Elev'6 5 Feet, Peak ev 7.25 Fee I
' . =
' :u -
15.4.1288 - Reflood Tr nstent, DECLG . - 0.8) 1. Mix, 10% S P
+ .
15.4.1 Refloo ransient, DEC (C. = 0.6) Imp Mix, 10% GTP 15.4. -90 Ref1 d Rate, 0ECLG . C. ='O.8) ! p Mix, 10% 5 TP 15.4.1-91 Re 'ood Rate,'OEC G .(C. = 0.6 Imp Mix, 10 'SGTP
- s. DECLG ( = 0.8) _ Imp ix, 10% SGT .
1 .4.1-92 ccumulato(F1 Accumulator lows, DECLG (C. - 0.6) 1 7p Mix, 10% TP - 15.4.1-93 15.4.1 4 28 Aluminum CorJosion in DBA Environment 15.4.1- 95, it Results'of Westinghouse Capsule Irradiated . Tests ,: - 15.4.1- 9623VolumeofHydr.ogenProducedFollowingaDesignBasisAccident ( ,, c.,o/ g ,' 15-18! ~ Oll3F/COC4
, ,, j
___ n I
- i
- SQN-6 i
LIST OF FIGURES (Continued) C Ngllgr, fM T1 tie 15.4.la97 N Hydrogen Concentration in Containment following a Design Basi: Accident ,7 15.4.1 98 tIContainment Nydrogen C'oncentration With One Electric Recombiner StartedQne.DayAfteraLoss-of-CoolantAccident 15.4.1 & Containment Hydrogen Concentration With Purging Started Nine Days After a loss-of-Coolant Accident 15.4.2-1' Variation of Reactivi,ty With Power at Constant Core Average .
. Temperature. '
~
15.4.2-2 Tralislent Response to Steam Line, Break Downstream of Flow > Measuring Nozzle With Safety injection and Off-Site Power (Case a)
* ,3.
- 15.4.2-3 Transient Reiponse to Steam Line Break at Exit of Steam Generator
~ #
- With Safety. Injection and Off-Site Power (Case b)-
15.4.2-4 Transient Response' to Steam Line Break Downstream of Flow, Measuring Nozzle With Safety Injection and Without Off; Site Power
~
. (Case c) . -
( . . 15.4.2-5 Transient Response-to Steam Line Break at Exit of Steam Generator ,.' With Safety injection and Without Off-Site Power (Case d) 15.4.2-6 UNI Flow Pat' tern Through the Vessel +
. . . s. . -
15.4.2-7 Integrated Flow of. B' orated Water versus' Time . u .i . Main feeditne Rupture Accident - Core Average Temperatufe 15.4.2r8 -
~
, Pressurizer' Pre.ssure and Water Volume as a function of Time. *
.+
- 15. 4. 4^-1 All Loops ~0perating,-%ne Locked Rotor P,ressure versus Time B 15.4.4-2 All .But One Loop Ophrating i Locked Rotor Pressure versus Time A11 Loops Operating,' One Locked Rotor Core Flow versus Time 15.4.4-3
- 15.4.4-4 'All Loops Operating 1 Locked Rotor Flux Transients versus Tjme Time 15.4.4-5 4
, All Rut One Loop Operating. One Locked Rotor Core Flow versus *
~
~*
15.4.4-6 AllBut0nei.oopOperating.OneLockedRotorHeatFlux'versusTime
*?
15.4.4-7 All Loops Operating, One Locked Rotor Time.'versus tlad ' Temperature 15.4.6-1 > Nuclear Power' transient BOL HFP' Rod Ejection Accident l
~ -' ,
s 0113F/COC4 j 15-19 ' l * , __ ] L._ .___ .__ _ __ . _. _ = _ . . _ . - . -m: r n ' -
q o x
~
l
..t ,
.SQN-6 '
x .,., thermal-hydraulic parameters, the. code accepts as input basic driving . C functions such as inlet 1 temperature, pressure, flow, boron concentration, control rod motion .and Others. .Various edits provide channelwise power, axial offset,- enthalpy.. volumetric surge. .pointwise power,. fuel , temperatures, and:so on. -
~
Tht? THINKl.E code is used to pfedict the kinetic behavior of a reactor for
~
transients which cause a major perturbation "In the spatial neutron flux ' l distribution. ,
- TWINKLE is further described in Reference 17.
4 15.1.9.8- WIT-6 ,
- y. s.
WIT-6 is' a one-region neutron ki$etics program with a single axial lump
' description of.. thermal kinetics making'it.useful in the analysis of .
transients in a heterogeneous, reactor core consisting of fuel rods, fuel rod cladding, and water moderator and-coolant. The code is basically a core model and therefore. generally uslifful for fast reactivity transients s which terminate befo're significant effects occur from the remainder of ,, j the plant, 1.e.,, transierts, shorter than the loop transit time. WIT-6 is used'in safety analysis.of. reactivity accidents from a3 .
.. subtritical condition.
WIT-6 is further described in,-Reference 18. - 15.1.9.9 sPHOENIX' .
~
The PHOENIX code calculates the 16dividual loop ~ flows, core (Iow and pump ' speeds as a ~ function Qf time subsequent to fa'llure.of any number of the . , , , reactor' coolant pumps. The analysis is based on a momentum balance
-around. each reacter coolant loop and across the reactor core. This.
- momentum tal~ance is combin'ed with the c'ontinuity equation, a pump momentum balance-and ~the pump characteristicsa Any number of reactor -
cool' ant loops are Rcommodated u_p to a' maximum of 6. PHOENIX ls further described in" Refer'e nte 19. 4 . 15.1.9.10~ THINC , The THINC code 4s described in Paragraph 4.4.~3.1. ,, 15.1.10 .' Upper Head' Injection b lOW
-nar uMZ Af Saw HA,s %eca RENove -
pressure lThe UHI . 5 tein shown scnemeticaifin igure 15JI.10-1 is a hi . or system, separate and,in pendent to the ECCS. whl provides accumul L. The addit! .al cooling water flow dir tly into the reactor ves ,. consists of two high pre ,ure accumulators operati at a nominal syst -1 and provides a volume f approximately 6 l pre sure defined in Table J.3 tank and 1800 cu. ft. of compressed 1 0 cu."ft. of liquid ~1n o itrogen gas-in the other. A membrane is provided i the gas crossover g. 15.'l-15 0114F/COC4 , e
.- .- . . . - .- - ~- - -
- < . j s ,
4 SQN-4 1
'+
be%W -
.line betwee the accumulators to s arate the water and gas
~
the line '\ -
' and thus' nialze the; amount'of n rogen gas which can diss ve .in the , ,
W rated emergency core- 11ng water from.the wat . filled
' (water. two 12. inch lines which ranch close to >
l *i 'accumul tortis delivered throu ' tthe r ctor vessel into four inch lines, and then'd! harges into the j l
" rea or, vessel head through he,four. head-pe'netration .
- d
(
,,r j .-
- 'I the; reactor' coolant p ssure drops'below the-n nal pressure in the ,
i line will tear open at *-1 as atcumulator, the
~
rane in the' gas crossov W^ about 40 psi.(AP) an low'will'be inttiated isolation valves ~an injection lines into the eactor vessel. Mechant
. operation of the s ng-disk check valves by differential pressure
~
th ugh the normally open 1 (4' I 1 the only action r quired to provide the in ction path ,from the j i accumulator to ' e-reactor head.. :- s , , Although the HI system has been. desig ed primarily to provide added ' injection f r:the large cold-leg brei , operation will also , large achieved y. for a wid range of accidents inclu ng,.small pr'imary brea and small s ondary breaks,'and'depre urizatten of tne react coolant. ' syste The UHI accumulators ha been'mbdeled to prov e.this ' addi onal{waterdeliveryfor;tesetransients.
, L 15.1.11 Loss Of One (Redundant) DC System : .
'f
/15.1.11.1- Identification of Causes .. ,
The plant DC System serves as a p,ower source 'for DC pump motors. . ('- controls, and instrumentation. A descriptiornof this system and 1ts U .1 ~
' redundant design are presented in Subsection 8.3.2. The loss of one DC iSystem will be. define'd for the purposes of-this analysis as the loss of
,. one battery;and one battery charger,. . .
- '15il.11.2 Analysis of Effects and Consecuences -
- As discussed:i,n. Subsection,8:3.2, the plant has be.en designed so that the 7
loss of one DC System wUl .not affect the safe oper'at1on of the plant.
~
~15.1.12 References f
- 1. Supplemental info.rmation on fuel design transmitted from R.
- Salvatori, Westi'nghouse NES, to D. Knufh, AEC, as attachments to .
letters NS-SL-518 (12/22/72), NS-SL-521 (1/4/73), NS-SL-524 (1/4/73) and NS-SL-543 (1/12/73,w-(Westinghouse NES Proprietary); and , supplemental information on fuel design transmitted from R.
~
Salvatori:.. Westinghouse NES', to D. Knuth, AEC, as attachments to ~ letters NS-SL-527-(1/2/73) and NS-SL-544 (1/12/73). 2.' J. J. DiNunno, et al ,. "Calculat-lon of Distance factors for Power ' '
?
and Test-Reactor Sites," TID-14844,. March 1962. ; NO e,
- r. _ . >
15.1-16 Oll4F/COC4 - l
$:I,l' 7 *
} ._ 'h
..* : \*
. v.
., i
. ... s
. .~;:.I 'lj ;
-.t.
w s q; ,
+
- s. 't ll .
~
y t. belete. - 'i>
,gy 1.:
. g. : . .
- 2.-
p ..
~
- t. . ~
~
t N " e'
~
J B _ h '~ p sgf r c . 3 ' 8.
- y' - 1 y
- f; V' .I l x,
> 4{- s: .
. p j p ;
~
l-
'l-I
\. , l. _ . - . a C d , ,ci g j.
! IE p g i *
'[
j/' 4 a E-y ..,. J: f. ([ - h-i
} .
.o-t: .
i .__ -a f
- l. W m. ; . . . . . E_ . , .
- R i
.g -
l' , j
/'
A-.{ l- , FJ g' .., t .- L:
.I -
W;: .3S[ ses y '
- b S " ~ "*= ,
g ual
- - w g .
?
,05t$ *-
c 0 A
. ' 5 " *_ b ) .i .
2E5 . y e ! _ m m =
.- w oo ,4
- n. **4
.j l^..- a. ..
W . s-A k
~
% ti _.
l ,
> . W
_. =.
. 9,'
a
- d. ~
e w ,--,t.
^'"
'f. .
. Q: _. ,
)~ .
-}w . :.
- + -
{.
'TAett'15.1.2-2-(Sheot 2).~
' .;b(Continued). s ..,
~
masenar er, tartrat wreame ama c_ . mi = - y >
*j *- -
~
- r. _
~'.'
gestifAL IIS$$
! . etACifVffY tetFFICitWF$
A55gura ' INEennt F9utt GUFPlit b ASSINES ~ ~ pgSttATORfGBEGAIGR'" +
. '4'
- C0pputta - ,TEMPtaAftfGE~ . DENSITY '
FAIK15 e. y CODES UTIL12EB marl wamicel - appperants
.~' '
. nasr3 CONDif!Ost II (Centlawed) , i -
~
(. i - 3 Less of Normal .. - . gtROUT , ; Feedwater . . get M 1 '
- ' 3577
'_ a Less of Off-51t'e BLROUT '
' e .1 Power to the
' : len 4 sin "
' 3423 .$--
Plant Amalliarles ' ee - " (Plant Blackout) , 3 e, l, t . tacessive Heat MMtytt i , Removal Ove te 3.43 9 and 3423-tower;(t.' Tcedwater System ; e l
) .
s-NW1 functions * . I + Escessive Lead L0Fitass - Iztrease 9 and 0.43 tower 3423-i, g
- Accidental Depres- I surization of the LOFTR48 '
(9
- Upper 3423 R actor Coolant - -
System Accidental Depres- . g , g ,9 pc. /.g i 6 '
~
Fonctlen of -pense. surination of the . Moderator-O Main Steam Systee ' tensity - ritical)
, see 5ebsectica i k
^~ 15.2.13
. (Figure 15.2.13-1)
Inidvertent Operatton LOFTRAff
- sf ECCI During Power Operation g tower I
3433 2 e l
+
I t . .;
#l ,
l k O g CD , e
~
.esswitsee 'i-3 o ,
4
.g
________ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _ _ _ - _ _ -- . , =-- ,_ .n.~ _ - .- ~. -__ . . ~. .~ . - , ,
l-__-_--_____--_ *
.- . n.,
i
.a 0I '
~
7 .,.
.,r- .-,n.
!s '
g ._ - ,
,~.c=-
. .[ ,
'~- ' ' ~
\'=
7 TABLE'95.1.2 2 ($heet 4).
- A, (Contladed)'. -
4 m
, $tseenay er INfifAL ffMERITfGNS AND CDNPtifft CSK1.* . . -
' *4
' r* .
~
REACitVffY COEFFICIENTS ' 3 INITIAL #55 TMERMat PeutR 9 8Put
" Al5UMED , - .
ASSUMES ~ - DEIDERATORHOSERATOR '
+
COMPUTER.
- 3-O TEMPERATURE DENStfY-
. l' FAUL15 CODES UTILI2EB e - f AU*F1 IAUan/cci 90PPLERf21 ~ #9Rft1 :- -
.i
~ , r p ppf,N CONDITIONIVkContinved) 81 .
g, Major secondary -MhlNEL.TyNC J. Function ofi .i ~ J _? j -f~ ' system pipe rap-e
-9' .-
! p' Hederator -
ture up to and 3 - p Density See (Critical) laciuding double. *
,. 15.4.2 (Figure ,
ended rupture - taupture of a Steam 15.4.2-1) g, Pipe) - 9 Steam Gevierator NA M M4 IIA Tabe Rupture " 3577 - Single' Reactor PHOENih.LOFTRAN 9 Coolant Pump Upper 2396 and 3423
~
THINC. FACTRAN 'i - - Lacked Rotor I.'I i. Fuel Handling NA HL M4 Accident . 3577: g-Rupture of a Con- TWINKLE. FACTRAN g -1 pcm/*F 99L -
' Constsent- 9 and 3423 trol Rod Mechanism LEOPARD -26 pcm/'F 90L with louer H using (RCCA Ejection)
IImit shown i Figure 15.1.6*1
- t Notes:
4 (1) Only one is used in an analysis i.e. either moderator temperature or moderator density coefficient. . t [ (2) Reference Figure 15.1.6-1 - 4 g' e e, e
. 9563F/ ESC 4+ ,, - -
s -
^ - - -.
g',
- . _ . 4 ~, - -
._.~ _. _ ,
~
,9~ _ _ _
= 4 - ' - :
m '
._ I i
SQN-5 .
}
i
- The steam release as a consequence of this' accident results in an initial
~
r - .j increase in steam' flow which decreases during the accident as the steam-
.I[i pressure falls.- Thi energy removal' from the Reactor Coolant System (RCS) '
causes a reduction of coolant temperature and pressure. In the presence l
- S of aLnegative moder,ator temperature coefficient, the cooldown results in
~
a reduction of core. shutdown margin. i The analysis is performed to demonstrate that the following criterion is satisfied: Assuming a stuck rod cluster control assembly and a single Jt t: = 7::r- M & dab oesw - failure in the Engineered Safet l
~
& tt:d ;td after reactor trip' for a steam release y Features @ r equivalent to the h sis e/,J '& c-spurious opening, with failure to close, of the largest of any s '
steam dump,- relief or safety valve. (rncr -_
^
~The following systems provide'the necessary protection against an ,
accidental depr_essurization of the main steam system. I '. Safety; Injection System actuation from any of the-following: ' Reva '*w
- a. Two-out-of-three low pressurizer pressure. 3,g ,
l.,
- b. High-differential pressure signals between steam lines. ~
- 2. The overpower reactor- trips (neutron flux and AT) and the reactor trip' occurring in conjuncti,on with receipt of the safety ~lnjection L
' signal. '
i 3. Redundant isolation of the main feedwater lines:' Sustained high - feedwater flow would'cause additional cooldown. Therefore, in, addltion t'o the normal control action which will close the main feediater' valves'followlig reactor trip, a safety' injection signal
. will rapidly close all eedwater control valves, trip the_ main _
(
.feeddatee pump, , and close e eedwater isolation valves.
l ' o peram. wr_p' A W O O V/%45 15.2.13.2 Analysis of Effects ana conseauences Py tn% Awa l {M
,N'ethod of. An'al ysi s ' y The following analyses of a secondary system steam release are performed l- for this section. .
brrod l y
- Reference code,
.1. A full p-lant-dl} ital computer simulation, "'T.Tt.
to determine RCS temperature and pressure ur ng ~
^
@^18 3nu%@su J.s mer
- t0r 2 0: ^^t return :riti 4 - 5 (
_2. An(ekoarane W I to + determine V that the The,following conditions are assumed to exist at the time of a secondary systerbreak accident. _
,e %
\
90 15.2-39 COC4/0115F O w e e' ,ts -
' s.
7... - j
,,/. 1 -4.-'-
m,,. ,. .
-- 4 .
p ~.
~
n . F'.
, . . - . ~
i -
. Insert'D- -
p
- a. .Two-ou_f-threilow steam line' pressure signals .in any one loop.-
i f' b. -Two-out-three low pressurizer pressure signals ~-
,. c..Two-out-threehighcontainment.pressurfsignals - C i m .
A S r e e ' { . k e** 4 e F e
* *= * *8.*- 4
- g, . . . . .. ,
'9 s
l K C , er
*w 9
e (' 9 4 ea , 9 (- e r . i
+
!- i
- 1.
- a:
kI'I ' p.3 he . .sd, + . }. .
- e *# #
l . 4 & 4 s'ke
"% e s# e 9 9
9
= - ee .
.A
% 9 9
- W"
, = m .
J. . ta r
'OI 0.
1+~e v - _ __ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ . _ _ _____ _ - _ _ _ . __ _ _ _ _ _ ._
7 SQN-5 j o I
- 1. End of ' life shutdown margin'at no load, equilibrium xenon 5 -
I
' conditions, and with the most reactive assembly stuck l'n its fully withdrawn posillog. - Operation of rod cluster control assembly banh ;
- during core burnup is restricted in'such a way that addition of - 4
. positive reactivity in a. Secondary system break accident will not
- lead to a more adverse. condition-than the case analyzed.
< 2. K. negative moderator coefficient corresponding to the end of life-rodded core with the most-reactive rod cluster control assembly in.
theifully. withdrawn position. The~ variation of the coefficient with temperature and: pressure is included. The Keff_.versus temperature
~
at 1000 psi corresponding to the negative moderator temperature m .
- - - coefficient!used plus the Doppler temperature effect, is shown in Figure-15.2.13-1.
3 ~ Minimum ceoability for injection of high concentration boric acid
' solution corresponding,to the'most restrictive single failure in the Safety Inj1ction. System. The injection curve assumed is shown in g _ Figure .1, 5 2.13-2. This corresponds to the flow delivered by one g6 :
i- charging pump del 1vering its full-contents to the cold leg header. ##N
- L No creait has Deen taken for Ine low concentration boric acto wntch LuA7- *
- must,be' swept.from the safety injection lines downstream of the ,
boron' injection tank isolation valves prior to the delivery of high
- g
. concentration. boric acid (20,000 ppm) to the reactor coolant loops. ,
l 4. The case studied is an.' initial total steam flow of 228 lbs/second at l-1015 psia from all s. team generators with offsite power available.
-Thi.s is the maximum capacity of any single steam dump or safety -
valve. Initial' hot shutdown conditions at tilne zero are assumed (' since this' represerts the most pessimistic initial condition. Should:the [eactorebe just-critical or operating at power at the time of..a steam release,.the reactor will be tripped by the normal overpower girotection when power level reaches a trip point.
. . Foflowing a trip at +ower the RCS contains more stored energy than at no-load, the average coolant temperature is higher than at no H load and there is.appractable energy stored in the fuel.
. 'Thus,-
additional stored energy: is removed via the cooldown caused' the steam-line break before the no load conditions of RCS '
-are reached. After the additional stored energy is removed,
~
5 cooIdown.groceeds in the same manner as in the analysis which
- assumes no load eqndit' ion,at time zero. However, sinc ~e the initial steam generator water inventory is greatest at no load, the -
L
- magnitude and duration of the RCS cooldown are less for steam line breaks occurring at power. .
l: . L 5. InNomputingthesteamflowtheHoodyCurveforfL/D-0isused.
" .w 6 y 15.2-40 C0C4/0115F .
I e , \[
- l
~
.- G., . . . .W * .
., . _ _ . . ~_ . _ . _ _
4 aus t .
.. i a , ;
y , r.
- a m -- '
- l l., :(
l e 1
~
.N .,
- .\
l 1 I y v ~
.. . ~ ~
~
8 The analysis conservatively assumes that the safety injection lines. ' downstream-of the'RWST contain no borated-water (0 ppm). This water must- . : be swept from the safety injection lines prior to the delivery of boric .1
" acid,(1950 ppm) from the RWST to the reactor coolant loops. (
i-
. )
e 9
'e
't p e...
- s. . .
g o ' * *
'# W.
e , , g ..*
" W 'Re%/*% %* ,, e t ,% ,J' ar
**y. ",
. 't .'~
i m ~.
+
% . e
./ . ~ *,,
- l. . ,:
s 013 , ~ , , e
'.._. ______m__U._._-_____._______.____--______ ~ --
, m- '
, l SQN
- q M
y _ 6.-; Perfect moisture- separation in the steam generator is. assumed.. e: Upper h d injectio ystem (UH is simulat . As sta d in- , I
,,'7.[WCAP-8185he signift UHI is to r tard the essure I
' t effect .
secreas of .the RCS. This in tu , reduces t flow of rated j
- water rom the Saf y injectio System. Th potential y the UNI detrimen 1 (p "
eff t is compen ted by bor ton provide
~
/.
.. Results --
l The results presented are a conservative indication of the events wifich would occur assuming a secondary system steam release since it is ' postulated that all of. the, conditions described above occur simul-taneously.- < [sfuu' , Figure 15.2.13-3 shows-the transients arising as the result of a steam ' tw7v release having an initial steam flow of 228 lbs/second at 1015 psia with 5"I steam release from one safety valve. The assumed steam release is the maximUni-capacity of any single steam dump or safety valve. -Safety g ' Injection.is conservatively assumed to be initiated by low pressurizer pressure although steam line differential pressure would provide a more
-rapid signal. Operation of only one centrifugal' charging pump is ; -
considered. Boron solution.at 20,000 ppm enters the RCS providing / sufficient negative reactivity to maintain the reactor well below criticality.., The reactivicy transient for the case shown in Figure ( 15.2.13-3 is more severe than Yhat of a failed steam generator safety or A
~ relief valve which'is terminated by steam line differential pressure or a ) -
falled condenser dump valve which is terminated by low pressurizer
. pressure and level. The: transient is quite conservative with respect t6 cooldown, sinc'e no cr,edtt is_taken for the energy stored in the system - t metal'other than that of the fuel elements or the energy stored in the
- other stelm generato'rs. Since the transient occurs over a period of aoove- five mbutes, the neglected store energy is likely.to have a {
significant effect in ' slowing the cooldown. =
. . 15.2.13.3 Conclusions
~
The analysts has stiown' that the criteria stated earlier in this section is satisfied. Since the reactor does not return to critical the possibility of a DNBR less than 1.30 does not exist.
* ~
r .
-15.2.14 Sou~rious Oderation 6f the Safety Injection System at Power
~
15.2.14.1 Identification of Causes and Accidents Descriptions Souridus SIS operation at , power could be caused by operator error or a f alse electr' ital actuating signal. A- spurious signal in any of the folloGing channels could cause this. ipcideg. ._ G
}
% M 15.2-41 t0C4/Oll5F 4
. ;f ' *c ' *,~ . s.: . . . ,
..?..,
i c, , v 3;.o - L
). a s e # T* .f
_6. Perfect'moisturi separation in the steam generato'r is-assumed.
. cit
~ 7 The auxiliaryifeedwater system providss'a flow rate of.WGPM-to each steam generator.. This aux 111ary feedwater is not required to mitigate the transient and-is modeled to increase the' severity of the core
'cooldown. . '
, Balul11 . _ ". s '
The calculated time sequence of events fort this accident is-listed in Table 15.21. The results presented are a conservative indication of the events which would occur assuming a secondary system steam release since it-is : postulated that all of, the conditions described above occur simultaneously.
- Figure 15.2.13-3 shows the transients arising as a result of a steam release ' * -
having an initial steam flow of 228 lb/second at 1015 psia with steam release - from one safety valve. The assumed steam release is the maximum capacity of any single steam dump or safety valve. Safety-injection is initiated charging pump-is assumed. Boron solution- Operation of only one centrifugal automatically by low pressurizer pressure.%(41950 ppm ente - sufficient negative reactivity to assure that the DNB design basis is met. The transient is quita.conservat.ive with respect to cooldown; since no credit is taken for'the energy' stored in the system metal other than that of the fuel ' elements or energy stored in the other steam generators. Since the transient-( , to have a significant ef.fect in slowing the cooldown. occurs over a period of 15.2.13.3:- Conclusient
'The analysis has shown that the criteria stated earlier in this section is
'satisYied. Since the minimum DNBR remains above the limiting value, no #
,;,,,. g ,, consequential damage to.the. core.or reactor system occurs.
, v. ,
~ .
4
- n. =
. m .
. a. : . .
~
f v .
~ - -
4 s '
. 014
, - 3 15.2-41 COC4/Oll5F -
{
3
~
L 8 SON-3 4 o . .
~
l t .- . TABLE 15.2-1 (Sheet 6) l7.~ , (Continued)- - P g -TIME SEQUENCE OF EVEitTS FOR _. i J E CONDITION II EVENTS
, ' Accident Event -Time (Sec.) -
,- - Excessive Load Increase - s .
g " l '. Manual. Reactor 10% step load increase 0 [' ' '
- . Control (BOL)
Equilibrium condittor,s reached _c' l (approximate times only) 200 <
;c 2. : Manual Reactor . :
Control (EOL) 10% step load increase 0 Equilibrium conditiens reached- ,. (approximate times only) 50
, 3. iAutomatic Reactor - ,
Control (BOL)
-, 10% step load increase O' 4 Equilibrium conditions reached (3)
- 4. Automatic Reactor' .
I- Control (EOL) , 10% step load increase ,0 --
; -- Equi' librium conditions reached .
(approximate times only) 50 l
'Acciden'tal deprlssurization -
l--- of the Reactor Coolant System Inadvertent Opening of one RCS'
>:.- Safety Valve , O l- .,
Reactortrip 29.3 s " __' ' Minimum ONBR occurs 31.5 .
~
Accidental depre.ssurl u ttorF , ggg ace of. the Main Ste'am aS'fefy System Inadvertent Opening of one
. main stea'm safety or relief *ff
. valve. -
O h.t.r#7h
- Pressurizer Empties 160 l ~ . .
1 . . V - 20,000 ppm bor-en reaches .'
'~
RCS-loops ; 197 UHI initiation time 221
- l. N (3) Did not reach equilibrium within the time scale of Figure 15.2.11-2 .
d 3 Revised by Amendment 3 C0C4/0723F
,2 .
0, 'i
*s .
u
-y e .
Insert h .
;['&
Accidental Depressurization. of the' Main Steam Safe.ty System Inadvertent opening of one
., main sfeam safety or relief {
- valve 0
;e ' ' '
Pressurizer empties - 221
- - , Low pressurife~r pressure
~ w -
safety injection setpoint reached 229-
. Feedwater isolation occurs 238' !
Safet'y injection occurs' ' 257
' ~
. Boron reaches the core' 268 Criticality attained does not occur
- . i c..
.a.
.; -e ,
n f s .
. ~
s y G. =t t i - r =
, u. ,
l~ 43* <
- - r o a .
.. =
.+ .
% e a #
,, , ~ .. -,
~.
ysy . N a w. 1
. A s
i
y,
- , .6 .
= -
50N-3 , Q c
' ~
TABLE 15.2-1 (Sheet 7)
- 1 -(Continued) -
,. TIME SEQUENCE OF EVERTS FOR _
?
[ CONDITION 11 EVENTS 3 Accident- Event Time (Sec.) y- -
- - Inadvertent. Operation of . - - ' \ '
; ECCS.during' Power Operation Charging pumps begin injecting 0 .g borated water Low pressure trip point reached 64 I Rods begin to drop 66 g e n IV ev I-Major Secondary' stem ,
yn ;4
.t
^
Pipe Rupture- , 1,. Case a Stea line ruptures 0 T ..
. Cr icality attaine 18 i ?-
P essurizer empty 15 J. ,
'- - ,000 ppm boro reaches loop 20' .~
. .UHI initiatio .ttm,e 16 2-C.
a .
. Case b Steam Itn ruptures -
'- - 1 Critica ty attained . 4 'M
' - Press izer empty 17 M
- c i
20, ppm boron r ches loops 21 ,
-f U initiation ti e 25.5 m.
. ., ~
~
Steam line ru ures 0
- 1. se c *
- m. '
Criticality ttained - '1 '
;- Pressurtz empty 6 _
. . 20,000, m boron reat s loops ~
30 4 l
' UHI 1 t1ation time 17 c .
l t . S am line ruptu s 0 - b 1. C ed .
, , .. riticality at ined 17
*~
- Pressurizer pty 18
~ -
- 20,000 ppIn oron reache loops 32 -
-
- UHI initia lon time 39
( ' - Revised by Amendment 3 ~ b i e ,
% g COC4/0723F 0
.a.
, , . . . . - - _ = - . . . -
)
.,s 6837-97 I
; M -
l , < l .06 ' .
~
(EN'# -
~
2ERO PowtR. 1000 PslA #
* - , END Of LIFE 80betD
~
1.05 - cott wiTu ont acca ! stucs FULL DUT 4 e 1.04- - , '.s -
= 1.03 -
5 s-
.u g ;s* -
w- . i,- -
' - l.02 .-
r .. .:: 3 . . .
- .g,
'. . C _
g
. ( l .0.l .
l: . . i.00- - i .- L
- 1- .
It. % 0.99 -
"'s ;*- 1
' l- 1 I 1 I -
. O.98 250. , 3b0 350 400 450 . 500 550
- 2 COREAVERAGETEMPERATURE('F)
~ ., .
4 ,* W * ~ _
~.
~ -
s ,
- gure 15.2.13-1 Variation of KEFF wie h Tegem I
. x -
t m
- e
'b '
q e C '*@ ; , 0 -
- y. _ . . . _ _ _ . . _ _ . . . . - . . . . . . _ _ _ - __
.i' ^
r :
-E
- 3. . . l
'4 80". .
I
;- WLTIPLICATl0ll FACTOR K i y.
4.- >
- emb q a g-
- N ~
l .. m .- t t g
, ]
4 U T e
* ,' .. , ,e
. A g es !
_ 9e , E
-3:
e r fe: - m Due In [ ;-
- e.
- W , ,s t.
w
,sg -
e jt ; ,4 e. . ,v v G.. *
- l. .
. R ,
' .3 e U * ' -
g
- e. *
. -e N.
s
<I a
% p $ O
.. .. M
. < e .
l- % - o l .. , .,~ ~ g c -
== g.D L . v ',
y I ' -
-m 4 O
4=== ,
., .s, .
c ; M-wt*
. .z .
- e .
'.~ i h0 g
- 9 a a-p
- A g
- 9 g .
T E
* == 1.
- h g peg egg em fut
# e y 1 :?
i , fue g -
. d. .
==
5
- s eG P 1- .
'4 m-gi 3 m put e en "of e * .
899
,d >k) "
, E *
^
m *a , e 4 5 9 9
~ , - - - _ _ . - - _ _ _ _ . _ . _ _ _ _ - - . - _ - . _ _ . . _ - _ - - . _ _ _ _ - . _ _ - . _ . _ . _ - - _ _ _ - - _ _ - _ _ _ _ _ _ _ _ . - _ -
y $ ,, 4
- I 2 ef, , n , of t
w m_ ' ' , ._.c A - v 74 y p..:
. . , , a.
)' ,
.- f- 7000 .. .,.
f _ N ,
..- r s 3
, .y
- .t, i
a , , gn .
~
d % i t
)
__// >- , I
.J..4, : , . . . .
. r .
4
-2000 - ' . - .
/ 5 --
^
, 1800 -> - -
.. j
~
t
- 1600 ' ---
. . a
=;
* . 1400 . - -
7 i.g - r
. ~ .
'"h l1200- .: - .
- 1 g .,. . .
P. y u
*W.lo00- -
l-L .- l . l , < . .800- . 1 N; ~~
, ,J s . i.
l ..; 600' -- s , 400 - l .. .
=
. - % .-- l 200 -
i .. .. i . ( O l l l l l' I I , s E 0' 100 200 - 300 400 500 600.. 700 800 , n " - SAFETY INJECTION Ft0W (GPN) .. 1 Safety injection Curve (EPLAL6 .
., - Figure 15.2.13.2 a l
? .
r- . i o ,
. . . .. . . , x.
. _ _ _ .A .
~.. a.% c.i. AmCW.-:~.:n~.*: .A.'
4
-. . .4 ,
.o , - c. -
-3 , ,e .
~ g 4
-: Figure 15.2.13-2 ' Safety Injection Curve
- . 1.
g_ 12 - 4 O- t. 5 1.6 - . i 'y 5 t - 1 3 ' , E. .1.4. - '. -
. ,i.
n . . 4 :;;
- dh.
1.2 - . sg w , .
. ua - - -
.g 1.0 - ,
<- c;. ii na i3 g '
IE Lo.a - - -
. 6
- r m
u.,.
~ gg . ,
../ .- e v .; \ a.g -
.g 10.4 - i .. .
't .
. N
. T g,1.- -
.', ; Q,Q , -
3 f ' . . -
-i
- , .i
+
. , ;4,e. ~
. y 0 -..
1001
-200- -
300 ,s, m
- jof@%
~
~
,. y
- .5AFETY IIMEQ FLOW (GPM)l - .. -
_..,.c . 8,0 -g- . .F. ; < . . . .;. . - 1~ - c 2.
*- ..r-.- a
- 1..,"~.
a n. ; ~ .. Q a 4 s +-[,
- 7. ;-
r' . *
.r : r - , .j,.- -'g.
+c -
'~ ,- -
- , ' . 'l i
- 4 h t4 e n + . . .
1
. <j -.
~
-~ 7635-36
\+. - .
pkA#-
.4000 ,
I ... _l I I -. PRESSUR12ft EMPTIES AT I00 SEC0ll0S-
+
20,000 PPet 80a0# REAClitt LOOPS AT 197 SECONOS- ' 3000- . .
.2 ~ .
a r s . w.
.s-
*'.2000-w '
.t , .
;, m li tcon .
- -i .
# 4 g ,
'O ,
A
... u. _ '600' .
o w 5% -
- i
-5w w~ - -
C. . g MO -
- o. . ..
.l -
g go_. _ N} *
-5 .
200 .[
=2.5' -
i gil 0 - y
. m .- ,
;- -2.5 -
o - 15 ' ' ..
-l
- i j .
l . l .
' G -5.0 . - .
0 - 100 200 L . 300 M0 500 + TIME (SECONDS) Figure 15.2.13-3 Transient Response for o Steam L.ine Break Equivalent to 228 Lbs/Sec et 1015 PSIA with Outside Power 1 Avelleble - N 1 i t .
.c . - . . a :4 : % .: m L D :3 %
!j i
.-[,
v
. f r. ?
.2500i[ .
**me ..
.Q 3'.'2000..
g 1500.- - t
~
- 1000. . <
- u n ' !
' ~
g 500., P
~0. '
O 100- 200 ' 300 '400 500 600 700 400- 900. s . L 600- - n m .
- l. -
'500 -
.w 400- -
E o . - 1-
- 5. 300. -
g, . ,
'E ^# .
g- - .. 1
=-
- 200 O 100 '200 300 400~ 500 600 700 400- 900 '
.c , .
~
=-
3000. ,, r a 2000.-
'E l- E 1000'T . ..
g -
.s .
0. l ,
. g -.10 00. -
c . . -
" ,-2000. .
. ?
-3000. ~~
s c. 100. 200. 300. 400. 500. 600. 700,
. 800. 900.
~ TIME (SEC)
Figure 15.2.13-3 Transient Response for a Steam Line Break Equivalent to "
-.228-tbs /Sec at 1015 PSIA with Outside Power Available d
0~
4 [ ', SQN , h . 1 - '
- ~
l
- 6. !Turb1'ne Load ~
. Turbine load was' assumed constant until the electro--
Then
. hydraulic governor drives the throttle valve wide open. -
turbint load 8rops as steam pressure drops.
~
7 ReactoruTrtp Reactor Trip was intt'iated by-low pressurizer pressure assumed at a conservatt,vely low value..of 1775 psia.. .. ..
- - J,
,. Results .-
~ .
, r' -
The trans16nt response is;shown'in' Figures 15.2.14-1 and 15.2.14--2. Nuclear power starts decreasing immediately due to boron injection but - steam flow does not decrease until'15 seconds into the transient when the turbine throttle valve goes wide open. The mismatch between load and nuclear power causes T.y. pressurizer water level, and pressurizer pressure to' drop. The low pressure trip set point is reached at 64 . secondstand' rods start moving into the core at 66 seconds.' , Af ter' trk,' pressures tnd temperatures slowly 'tise since the turbine is ' tripped and the reactor is producing some power due to delayed neutron fissions and decay heat. -
, . F~ -
, '_ ', y
, 15.2.14.3 Conclusions Resultsiftheanalysisshowthatspurioussafetyinjectionwithor ,
without immediate reactor trip presents no hazard to the integrity.of the Reactor Coolant System. -.
~ , - . , . ,
( DNB ratto is never less than the initial value. -Thus there will be no {} cladding damage-and no release of fissionjroducts to the reactor coolant
- (- L system.
~
Ifthereactor,doesnotthipimmediately,thelowpressurereactortrip
~
.J.
will;be actuated. This, trips the turbine and' prevents excess cooldown-
' thereby expediting recovery from the -incident. . .
. : n. ' ' '
~ ~
15.2.15- Refer'ences a
~
li . 1l h
- 1. N'2.~C.Gangloff,'AnE'valuati.ondf' Anti'cipatedOperationalTransi'ents
;in Hestinghouse Pressurized, Hater,R,eactors " HCAP-7486, May 1971..
- 2. D.-B. Fairbrother, H. .G. Hargrove, " HIT-6 Reactor Transient Analysis _
- . Computer Program Description," WCAP-7980, November 1972. ..
. 3. .C. Hunin,'"FACTRAN,, A Fortran Code for Thermal Transients in UO, -~ '
Fuel Rod,"'HCAP-7908, June 1972.
.D -
- 4. l T..H. T.. Burnett, C. J. McIntyre. .J. C. Buker , R. P. Rose, "LOFTRAN E . Code Description," HCAP-7907, June 1972.
~
Y %:.- - . ( how civ;" TY. :. e + . bo F7AA Al Cooe- pekasig,Q " ' . WCAP -770 7- P-A [i siosierssy ) WMP- 7 fo 7- A - 00^'*~ 0*Nie7#k# han. /9fy' 15.2344 *COC4/Oll5F .
~
1 i,
-- -_-_ _ -__ _ _ _ _ _ _ [_
7 y, d 1 m '
, . i
- SQN-1
~
15.3 CO WITION III - INFREQUENT. FAULTS -
- l 1
By definition Condition III occurrences are faults which may occur very infrequently during the life of the plant. They will be accommodated J with the fathere of only a seafl fraction of the fuel rods although j sufficient fuel damage might. occur to preclude resumption of the ! operation for a considerable outage time. The release of radioactivity ' wtll-not be' sufficient to interrupt or restrict public use of those areas beyond- the exclusion radius. A Condition III fault will not; by itself. - generate a Condition IV fault or result in a consequential loss of function of;the Reactor Coolant Sistem or containment barriers. For the
. purposes of this repor.t the:following~ faults have been grouped into this '
category:.
. A
- 1. Loss of Reactor Coolant, from Small Ruptured Pipes or from Cracks in Large Pipes, which actuates Emergency Core Cooling.
.D
- 2. Minor. Secondary System Pipe Break. - M
- 3. In1rdvertent-Loading'of a Fuel Assembly into an Improper Position.
. ~. .
- 4. Complete Loss o? Forced Reactor Coolant Flow. .-
,- - - 5. Waste' Gas-Decay Tank-Rupture. -
- 6. ~ Sin'gle Rod Cluster Control Assembly A Yhdrawal at Full Power.
' ~
The time sequence of event's' during appilcable Condition III faults 1 and
~
L . 4 above is shown in Tables 15.3.1-1 and 15.3.4-i. ' I - ! l . 15.3.1: 1.oss of Reactor Coolant From Small Ruotured Ploes or From Cracks '. ? in Larae'PToes which Actt stes Emeraency Core Coolino System
,u
'. 15.3.1;1 Identificat. ton.of Causes and Accident Description . ' -
\ . . .-
' A loss:of coolant accident (LOCA) Js defined as a rupture of the' Reactor Coolant-System (RCS) piping or of any line connected to the system. See l Section 3.6 for-a m6re detalled description of the loss of reactor coolant' accident boundary limits. ' Ruptures of small cross section will
- =
cause expulsion of the coolant at a rate which can be accommodated by the.
, _' charging pumps which would maintain an operational ~ water level in the -
pressurizer. permitting the_ operator to execute an orderly shutdown. The coolant whi,ch would be , released to the containment contains the fissioni l products existing in it. ~ [. The maximum break size.for which the normal makeup system can mai.nta'in- .; the pressurizer level is.obtained by comparing the calculated flow from .. the RCS through the postulated break against the charging pump makeup
~
- f. low at normal-RCS pressure,it.e., 2250 psia. A Makeup flow rate from
; one centrifugal charging pump is typically adequate to sustain .
pressurizer level at 2250 psia for a break through a .375 inch diameter hole. This break results in a loss of approximately 17.25 lb/sec. r -. - 4 15.3-1 0116F/COC4 -
. . i y ..,.
l l Should a'larfrrlreak occur, depressurization of the RCS causes fluid to I flow to the RCS from the pressurizer resulting in a. pressure and-level j 6ecrease in the pressurizer. Reactor trip. occurs when the pressurizer ' , low pressure trip setpoint. ls- reached. The Safety Injection System is L actuated when the appropriate getpoint is reached. The consequences of L the accident We limited in-two' ways: .
- 1.. Reactor trip andJorated water injection complement void formation Lin causing rapid reduction of nuclear power to a residual level ,
corresponding to the delayed fission and fission product decay. 2... -Injection of borated water en'sures sufficient fIooding of the_ core _ .
, to' pre' vent excessive clad' temperatures. .
Before the break occufs the plant ts in an equilibrium condition, i.e., ' the heat generatid-in the core .is' being removed via' the secondary system. During blowdown,' heat from decay, hot internals and the vessels continues to be transferred to the RCSr-The heat transfer between the ' RCS and the secondary system may be in either direction depending on tfie-relativt temperatures. In the case of continued heat addition to the secondary, system' pressure increases and steam dump may occur. Make-up to 1 the secondary side is automatically provided by~' the auxiliary feed'ater -' ' w pumps.- t
~
t V
. The safety injection signal stops normal feedwater' flow by closing the .
- main feedwater line Isolation vaTves and IrtJtiates emergency feedwater
_ flow by starting auxiliary. feed W. ^atry flew = 'd; '- ()' - ? p i sure;
....."....su...; , tr.;......"C d;;; ;;;r '._- .:i . t:.."u.. : r- c-Qt g% -
_.2 ...._..i. .. .. .
. u..a .... ..
. "":_I". C Ila' :sR a li ' '2Z;'.;"'.'EC '..PZ.':'i. 7 ~i "' ~ ~ -
7'f"'l'a'CTr " Z.!~ . s il' 'T..'.'" Ti.1'II'i.Ifi'.' "'J'; . . i t i. m u 7
. '."' . ' E T '. . '.".,.R m. . , .' Z. .-'.A. :. '.. .Z z _ : i ,u '. ~ - ..:
'.'. .' .'.. '.. 2 . .2. !. . .".
When th RCSdepressurize[toapproxima sia the cold leg . L, - accumulfitors' begin= totnject water into t'he .n,.;w coolant loops. The L reactor coolant pumps are asstreed E5 be tripped at the initialization of the accident and effects .of pump coastdown are included in the blowdown analyses., .. , 15.3.1.2 Analysis of Effects and Consecuences
.,- gwse .
- Method of Analysis 14oTRumPCIS For breaks less than;1.0 ft' th "J LA';;M i > gital computer code is ~
employed to calculate the transient aepressurization of the RCS as well
-+
as to describe the mass and enthalpy, of flow through the break. , , 15.3.1.3 Small Break LOCA Analys Usino WFLA 869 ThdNFLAIHp gra used n th anal is of the ynaWti j%akdoss of
' /
goolant ac en is a exte on thetFLASH AC2) co6e'dev, eeloped the -
/ .
.. ~
15.3-2 0116F/COC4
wp .>
]
n y . ' I [ SQN g.] c . C Reactor Coolant System P1De' Breaks
~
3
}
This section presents results of'the., limiting break size in terms of l highest peak clad temperature. 'The worst break size (small break) is .be i
-inch diameter break. The depressurization transient for this break is shown in Figwe 15.3.1-2. Thee. extent to which tne core is_ uncovered is 4
'shown in Figure 15.3.1-3
.During the earlier part of the.small break transient, the effect of the break flow is not strong enough to overcome the flow maintained by the
; reactor coolant. pumps through the core ~as they are-coasting down i following reactor trip.: Therefore, upward flow through the core is
, maintained. ,
.c:
'AlowRC$pressureRCPtripcriterionhasbeenincorporatedintothe 4 ?
emergency procedures to provide an indication to the operator to trip the' R(.. s- for small break LOCAs but< would not Indicate a need to trip the RCPs for-the more likely non-LOCA transients-and accidents where continued RCP f operation is, desirable. . s The:rettaltant heat transfer cools the fuel rod and. clad to very near the coolant. temperatures'as long as the core remains covered by a two phase ' E8aN *N sixture. _
- Twssar a 2,
-The maximuni hot spot clad temperature calculated during the transient is
( (,- 1485'F includina the-effects of fuel densffication as described b
~
Reference 3.fThe peak clad Temperature tr:nsient is shown in Figure '
- 15.3.1-4 for the-worst peak size i.e. -the break with the highest peak -
-clad temperature.' The steam flow r. ate for the worst break is shown on Figure 15.3.1-5. When the mixture level' drops.below the top of the core, --
l- ' 2 tnt steam flow computed i LASH provides cooling to the upper portion ' , of the core. -The roc film coe, cients for this phase of the transient
'are given in figu k 15.3.1-6. The hot spot fluid temperature for the' ..
W worst 4reak is .shown in. Figure 15.3.1-7.
' SE W *7V
! ak S1zT
, #Jl' p _v
.- Tw pr / e A spectrum of break sizes in. addition to the limiting break was analyztd. These additional breaks were the 4.0 inch, 6.0 indh and 0.5 i1 .- ft' breaks. Figuies 15.3.1-10 through T5.3.1-12 present the'RCS .
*~
L pressure transient. Figures 15.3.1-13 through. 15.3.1-15 presents the - core mixture hetyht trartsient for all the breaks.:The peak clad temperatures for all ,the cases are less than the peak clad temperature of. the 8.0 inch break. The peak' temperatures for the 4.0 and 6.0 inch and E I the 0.5 ft'.breakt- are presented in Figures 15.3.1.-16, 15.3.1-17, and ~ - w .
~
It should be noted that all small break sizes presented here result in ~ calculated peak clad temperatures much~Iess than those calculated for . l large breaks (Section 15.4.1). -
-l?gutd MW $v3ef7' W ;
15.3-4 0116F/COC4 ,. __l.________.__.____________________________'___ . - _ _
[ , 1 e%' . a
.i
~
I.NSERT fl - Page 15.3-2 and 3 "1 j I
'[- Replace sub-section "Small Break loCA Analysis Using WFLASH" of
, \, section 15.3.1.i with the following: {
. : . i 15.3.1.3 sam 11 Break inca Analysis timina NOTRUMP The:NOTRUMP'computercodIis-usedintheanalysisof-1.oss-of-coolant accidents-due to small breaks in the reactor ..
. coolant system. ,The NOTRUMP computer code is a state-of-the-art,
.one-dimensional, general network code _ consisting.of a number of advanced features.' A'mong these features.are the calculation of
, I
' thermal non-equilibrium in.all fluid volumes, flow regine-dependent drift' flux calculations with counter current flooding limitations,f mi'ixture level tracking logic in multiple stacked' fluid nodesi and. regime dependent. heat transfer -
i correlations , sThe NOTRUMP small break LOCA emergency core . cooling system (ECCS)-evaluation model was developed to determine . l the-RCS response _to design basie.small break LOCAs and to address
' the NRC concerns expressed . in' NUREG-0611, " Generic Evaluation of M Teedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse pesigned operating Plants".
In NOTRUMP, 'the' RCS is nodalized_into volumes interconnected by
, flow paths. The' broken. loop is modeled explicitly with-the
-intact loops lumped into ,a.second loop. .The transient behavior
'[ of the system is determiped from'the governing conservation
( equations-of, mass,-energy, and:momGntum applied throughout the system. .A detailed description of NOTRUMP is given in References 1 and'2. -
~
* .. n . ,
Thefuse'of NOTRUMP in the analysis involves the representation of the reactor core as heated control volumes with an associated- .- bubble rise ,modh to permit a transient mixture height calculation. : The multinode capability of the program enables an
'expligit_and detailed spatial representation of various system components. In particular,-it enables a proper calculation of the behavior of the. loop seal during a loss of coolant transient. .
~
Cladding thermal analyses are performed with the I4CTA-IV (Reference 11) code.whichusestheRCSpressure,'fuelrodpower[ , history, steam flow past the uncovered part of the core, and.
^
mixture as input. height ' history from the NOTRUMP hydraulic calculations, e s
~
007
--e - - - _ . - - _ _ _ _ . _- - ___ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
s .; INSERT f1 (cont'd)g **' LThe small; break analysis was performed with the approved
- (G. .
Westinghouse"ECCS Small Break-Evaluation Model [1
-c. .d E.
l p, Safety -inj ection - (SI) flow rate to the RCS as a f on of the E 4 system pressure is usod as part of the input and shown in Figure
- l. 15.3.1-1.. This figure refrasents injection flow from,the ECCS j pumps based ~on a composite of 25% degraded TVA performance curves and 5%-degraded test data curves. SI delivery to the RCS cold .i l
' legs is assumed- Q seconds after the pressuriser low pressure safety-injection-set point is reached.: RHR pump flow is not a !
' factor in the analysis since pressure remains above the RHR shutoff head during that portion of the transient considered here. Also, minimum safeguards ECCS capability.and operability have been assumed in>these analyses including; conservative assum line.ptions with regard " to spilling of ECCS water from the broken Hydraulic transient. analyses are' performed with the NOTRUMP code .
which determines the RCS pressiur'e, fuel rod. power history, steam ? L flow past:the history 2 Theuncovered part of the core, and mixture height core thermal transient is performed with the ' g LOCTA-IV. ' ' code-operating at 102% of . . licensed Both calculations core, power. assume the core is I e
- g @
- 4 s,
* ~
. .D '
- . a. ,
[: , - :-
~
e * *
' GE quum O
a
~
l l' W u o
% e
. a 9
0
~
008 ' W
L J
'! froi Y son.g a p
- y .{fyl fysd fh. f Westinghou boratory. The rogram permits a detailed spatial representation of the RCS. -
The RCS is. nod'alized into volumes interconnected by flow-paths. The' [ broken' loop is modeled explicktly 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 epplied , _ l .throughout the system. A detailed description of NFLASH is-given in j i
-Reference 1.. ,
[- L The use of.WFLASH in the analysis, involves, among other things, the representation of the reactor-core as a heated control volume with the / associated bubble rise moder to' permit a transient mixture height .I calculation. 'The multinode capability of the program enables an explicit ' t and detailed spAttal~, representation of various system components. In particular it enables a proper calculation of the behavior of the loop sealduringaloss-of-coolanttransieng Safety Injection flow rate to the RCS as a function of the system pressuts is used as part of the input. The Safety Injection (SI) system was assumed to b~e delivering _ to the RCS 25 seconds after the generation of a safety injection signal. '
~
h Tor thessianalyses, the SI delivery considers pumped injection flow which a ' '- is depicted in Figure 15.3.1-1 as a function of.RCS pressure. This .
, figure represents injection flow from the JCCS -pumps ba' sed on -
. performances curve's degraded 5 percent from the design' head. The 25
' seconds delay includes time required for diesel s.tartup and loading of the-ECCS-pumps onto the-emergency b'uses. The- effect of RHR pump flow is not considered here since their ' shutoff head is lower than~ RCS pressure ' -
~
during;the time portion'of~the transient is considered here. Also-utnimum Safeguards Emergency Core Cooling System capability and operab,111ty has .been assurped in these analyses.~ H [ ' Hydratilic transient wakyses'are'perfernied with the NFLASH code which determines the RCS pressure,-fue17od power history. steam flow past the uncovered part of the core.and mixture height history. The hydraulic L k analysis is erformed using.102% of the Engineering Safeguards Design Rating (ESDR) power. 'The core thermal transient is performed with the - __ LOCTA This analysis uses 102% of the NSSS Core licensed
- 6 .
] g r. IVC 11] code. _
e = .
~
Refuits ' -- - hecoun ng for the'un drtajnti of the/ upper ead'irgection ccumula:6r set nts 11 ted in helargebreakJCCSL anal it s, Sec ton 15. ).1 FSAR. this a lysis w t perfo med ust g the et'poin that of t res ted I the mum cal lated ot spo clad emperatu e. Th set
~po t was t 130 sia wit a wate volum deliv ry of 710 ft'.
/ / y -
s
~ k
'._ te . ,
d- , f 15.3-3 0116F/C,0C4, , L + _
, n , ,. - e ;i: '- > ~" ~ ~~ ~~^ ~ " ' ~ ~ ~~'
. ll l l . 34 _iy l + '
-f3 '?
' ~~~ R yj; ': ,y.. .
k 6 a , ,
- n.o,
}
w .
^
+
.> a: .. ^; i p ,
, -R .
ts c. l-
$' INSERT #2 -rPageJ1'5.3-4l, Revise as follows:. '
I ., ] 3; " - ' f.y , t
,The maxian2m hot' spot. clad temperature calculated during the- ,
i 1 transientils.2105.5'F. n s ,
'l J ' ~ '
~ i v; 4 9" ' -
. 1 . .
a' INSIRT 13 - Page~15'.3-4,' Revise as-follows: . i
, , ...the steam flow computed in~NOTRUMP...
y ,
+
Ay *
,,. INSERT #,1 - Page 15.3 L 4) Revise as follows: - * ? ,
[
- r. ,
EAdditional? Break' sizes- . O s r s . . R A '.speltrum of- breaks in. addition.to the limiting break was j
, = analyzed. These additional breaks wer_e.'the 2.0-inch"a'n'd 4.,D' inch [
[ --c - brea)ds..jyigures 15.3.l'-10 and 15.3'.'l-11 present the R'CS pressure ) [
'translents:and Figures 1,5.3.1-12 a'nd 15.3.1'-13 prese'nt the core , a.
j
.y ;aixtuie height: transients for these breaks. 'The peak clad j
! temperature fof these casestare~1ess than the peak' clad
~
h temperature' f or; the '3 inch -break. The peak clad temperatuies for
~
- -- ~
. the"2.0 inch.and-4'0: inch breaks are presented in Figures
, 3.
L > , ._ _
'15. 3.J-14 and - 15. 3.'l-18, respecti,vely. ,, - -
,, , .3+ ~
s
; 1
', INSERT #5 - Page 15.3-4', R.eplace with the-following:
+
It should.be noted'that the 3 inch break resulted in a peak, clad w. temperature greater than the large break calculated with the 1981
-Evaluation Model with BASH in Section 15.4.1. This is because the small break parameters'inclu'ded such conservatisms as an increased.safetytinjection d'elay time of 47 ' seconds," Fq=2.7, and
'l
' ~
a 15% steam generator tube plugg3ng level. ~ 011
- -- w-y-w-
e ;
~
1 l
... I sm r, . . -
l
.) ,
\
15.3.1.4 ' Conclusions - Thermal Analysis ) For cases considered, the Emergency Core Cooling System will meet the AcceptanceC,riteriaasprosegtedin10CFR50.46. That is:. .
-l ;
. 1., The calculated peak fuel element clad temperature s margin to '
the requirement of 2200'F, based on an F, value o g.
- 1. The amount of fuel element cladding that reacts chemically with mater or. steam does not exceed I percent of the total amount of
~
;Zircaloy in the reactor. ,
. s .
- - , l
- 3. The clad temperature transient is terminated at a time when the core s . geometry is still amenable to cooling. The clad oxidation limits of I !
l 17 percerit of the clad thickness are not exceeded during or after E - quenching.
' 4. The core temperature is reduced and decay heat is removed for an M extended period of time, as required by the long-lived radioactivity Temaining in the core. ,
j The time sequence of events is shown 4n Table-15.3.1-1. Table l 5.'3.1 ~ 2,. l f presents _the results.of these analyses. N 15.3.2 Minor secondarv system Ploe Breaks - t 15.3.2.1 'Identificat' ton' of causes and Accident Desertotton , , , Included in4his grouping are ruptures of secondary system lines which . _ would result in steam release rates equivalent to a 6 inch diameter break
. . or. smaller. .,
p l 15.3'2.2 Analysts' of Effects and Consecuences. J. .. .
- Minor secondary system pipt breeEs must be ~ accommodated with the fallure
- of only a sull fraction of the.Tuel elements in the reactor. Since the results of analysis presented in Subsection 15.4.2 for a major secondary
- ~
systea pipe rupture also meet this cri.teria, separate analysis for minor secondary system pipe breaks is not required. I
- The analysis of'.'the more probable accidental opding of a secondary ~. -
system steam dump, re-11ef or. safety valve is presented in Subsection . 15.2.13.; These analyses'are 111ustrative of a pipe break rwivalent jn size to a singleNalve opening. . 15.3.2.3 Conclusions
~
The analysts presented in Subsection 15.4.2 demonstrate that the " 4 consequences of a minor secondary system pipe break are acceptable since a DNBR of less than 1.3 does not occur even for a mort critical major ~ g secondary system pipe break. r .. ,.
- r. .. '
4: s lS.3 S' - D ,U,C * -
- . ~ . - - - - - . . .. - . - . .
J l
, , , $QN .
- 2. If the reactor is in automatic control mode, withdrawal of a single C' .
rod cluster control assembly will result in the tumobility of the *'
)
other rod' cluster control assemblies in the controlling bank. The j transtant will then proceed in the'same manner as Case 1 described 4 above.- For such cases 44'above a trip will ultimately ensue, t 41thovyh not suff1ctently fast in all cases to prevent a minimum DN8 i
'- ratto ' n the core of less than 1.30. .. !
15.3.6.3 Conclusions . f For the case of one rod cluster control assembly fully withdrawn, with ! the reactor in the automatie or the manual control mode and 1n1t1 ally ; operating at full power wi n Bank D at the insertion listt, an upper bound of the number..of fuel. rods expertencing DNBR 1.3 is 5 percent of , , the total fuel. rods in the core. - I
~
j For both cases discussed, the indicators and alarms mentioned would
- function to alert the operator to the' malfunction before DN8 could "
s . occur. For cass 2 discussed above, the insertion limit alarms (low and ' low-Iow alarms) would also serve in this regard. 15.3.7 References _ U [%$,
- 1. V. J. Espostto, K. Kesavan, B. A. Maul, "WFMSN -4 FORTRAN- IV - N i
[ Computer Program for Sim014tton of Transients in a Multi-Loop PWR," ! ( -
, WCAP-8200 Rey 2 (Proprietary). NCAP 4261 Rev. I INon-Proprietary), '
August 1974.
~ -
2.. Porsching T. A.. Nurphy, J. M., Redfield, J. A., and Davis, V. C.,
" FLASH-4: A Fully Implicit FORTRAN-IV Program for the Digital Simulation of Transients,in a Reactor Plant," NAPD-TM-841 Settis r E__-'
Atomic Pcwer% aborator.y (March u1969 r 8 3 #
. W. A. Bezalla, C. L. Caso, A. C. Sp'encer, "LOCTRA-R2 Program ,
' Loss-of-Coolant Transtent Analysis," WCAP 7835, January 1972. .
~
S. Altomare and R. F. Sar'ry, "The TURTLE 24.0 Diffusion Depletion f Code." NCAP-7758, September 1971. ,
.g R. F. Barry, "L'E0 PARD A Spectrum Dependent Non-Spatial Depletion ~
Code'for the. IBM-7094," WCAP 3269-26 September 1963.
~
g[ F. M. Bordelon, " Calculation.of Flow Coastdown After Loss of Reactor Coolant Pump (PHOENIX Code)." NCAP 7969, September 1972. ' -
*/
f 7,4 T. W. T. .Burnett, C. J. McIntyre, J. C. Buker, R. P. Rose, "LOFTRAN Code Description," WCAP 7907, June 1972. g# C. Munin, "FACTRAN, A Fortran-!V Code for Thermal Transients in
.U0: Fuel Rods," WCAP 7908, June 1972.
~
- i
- r. . . y -
fg/tjf ' I C. 3' I!- . ..:*s'
.w- , .
e l l f - ~ ~
' I k IMEIRT #6 - Page-15.3-12:
. . . 1 Replace-existing references with the followings l
~
]
1.,. Meyer, P. E., "NOTRUMP, A Modal
- Transient small Break and ;
1 General Network Code", WCAP-10080-A, August 1985. . l
- 2. Lee, H., Rupprecht, S.D., Schwartz, W.R., Tauche, W.D., l
" Westinghouse Small Brqak'ECCs Evaluation Model Using the l NOTRUMP Code", WCAP-10081-A, August 1985. , , l i
l
.P
_ M i
- i 4 .
C . e I
. ~ ~
l . l I (. i m ..* g
.. g e
[. .
.f <
g * . P * , l a_ ' i . . , l . .
~
4.*
*9 7
g e .* l . l e *
~
- r. ..
014 . s
'O . p gp
t i
,f:-
- # i dei'W soM-s i
- r. . - F (l d ,10. M. T Youn , g- 1., sti se rg cy S ;
E Datt Mode ppit tion o P1 ts E 1pped lth a , I ecti . -8479 v. (Pr .teta ), -8 . ( -Pr rieta ), Ja ry 975. , =- # f F. M.4 ordelon, t', ', 'LOCTA-!V Program: Loss-of-Coolant !
- 7. , M Transient Analysts. P-8305. (Non-Proprietary). NCAP-8301
; (Proprietary), June,1974. -
pH . s l
. . :- , 1
*\ .
8 e g . 1 \m
- , t Ah
, j pq .
g e * .
) -
A s, . s - -
*g %%
. ; a.
- e 9 8 to ,
eO'
~~'
4
- e f.
g .W.
, Om
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O b i . . . - 1 . . l . \ * (
.I\) -
F e== lx 3 I3 I 4e
~
~ '
t , e e . . - 4 pg . l -
*
- I'TABIX 15.3.1-1 i
.g
getAe.e r
- 4 Imet SMALL BREAK , -
- g
~ -
' ~ '
8 4 TIME SEQUEIOCE OF,EVElrI5 , (
,i Case AnalysIa r ( . .
4.0 Inch
~
6.0 Inch 8.0 Inc t s . . , I , , ~
- Start .
0.0 0.0 ' O.0.. - l . ; . i
~
Reactor Trip signal Se' ram Time ' 16.98 , , 10.7. 9.65 - d
' I.
Top of Core lineovered 738.10 253.2 146.90 , UNI Accismulator Injection Begine 42.10 g'29.34 , ' ' , - 25.00 . , p. - I . i Cold Leg Accumulator Injection Begins .. 1438.6? " 448.4 ' 217.00 t . PCT Occurs 972.0
. c , 501.3 241.9 :
e .
- Top of Core Cow red 998.4 508.4 255.5 ,
e
' ~
~ \
t l - . I . .
- i. . -
+
ei ' Revised by Amendment 1
, 0573F/COC4
( o .- '
~
e ,
. ,-,
- ro n - - . r m,w,--. . - . - - -n-- ~ ~ - * - m-+ - ~ + - - - - - - - - - - - - - - - " - - - ' - - - - - - - - - - - - - - - - - - ~
, ,. .r -
~
D
;, - :ggg_1
'3 * *
.' t TABLE 15.3.1-1.
~
4 IDOP SMALL BREAK . U ,. , - t,; . TIME SEQUENCE OF EVENTS ,- 7*y, .. ". Case Analysis < '4 . . - ,. I l -
, . ::3. .2.0 Inch . 3.0 Inch 4.0 Inch i '. i l ' '
p Start - r 8 g - 0. 0 ;, 0.0. 0.0
..i g Reactor Trip Signal Scram Time' I
44 03 ei
- 18.28 10'.51 -
*4 Safety Injection Signal 58.'00 .
26.35 , 15.48 , ! Safety Injection Delivery' 105.06 ,, 73.35 62.48 j Top of Core Uncovered 2908.3 922.4 ..' . ._ 659.1 ' - Cold Leg Accumulator Injection Begins N/A 1691.7 855.5 j PCT Occurs '! ' J 3381.3 1841.9 1014.6 1
- Top of Core Covered .;- 3395.3 (1) , (1) g .
j 1. ; -
. . 4
- 3 2
(1) Transient was terminated prior to core recovery. PCT turnaround had occurred
- safety injection exceeded break flow, and the core mixture level was increasin,g. t
? . A t . s i ! g t
~
'I
' o ..
! - p : .
.J . . .
a g h
- _ _ . - . ~ - , _ - - .._.r , . -
., .~,. .. , m . - - __ .. . _ _ _ - _ _ . _ _ _ _ . _ . .
e . ., n - q ' ,.' ( / - w *~ %.- n .s ., a . sqw_1 t , U 't. ,. TABI 15.3.1-1 EPikcf 8 4 LOOP SM4LL BREAK ? f
- .4 -
1 IaseAnalysis
- 4.0 Inch 6.0 Inch 8.0 Inc.' 0.5 ft' #
i Results * , . , 3 . I e J $[ Peak Clad Temp.'(F) # % 1061.0 V75.0 11.5-
-1486.0 10.5 1240.8 10.75' Feak Clad Location (Ft.) " P; 12.Os . i Local 2r/B 0 Run (max)% , 0.374 0.5'26 0.532 ' O.374 i Iscal Ir/Na0 location (Ft.) , . 12.0 11.25 10.75 10.75 '
, I
., Total Zr/H 0 Run % r
<0.3 <0.3 <0.3 40.3 Hot Rod Eurst Time (sec) N/A 4b.8 241.9 166.0
' 'l Hot Rod Burst I4 cation (Ft.) N/A 11.25 10.75 10.75 -
~ ,
{ .
.. Calculational .J)ssumptions ,,, . ,
NSSS Power (Nydraulic analysis) MWT 1021mf 3570 Core Power (Rod Heatup calculation) 102% of 3411 f 1 Peak Linear 18owJr kw/ft 102% of 11.97
' Peaking Factor (At License Rating) 2.32 ,
Cold Leg Accianulator Water Volume 1050 Ft' (nominal) t ..
. UNI Accumulator, Water Volume Delivered g 710 Ft' (nominal)
Revised.by Amendment 1 I
, i
~- !
~
O 4 l .
. , 0574F/COC4
- G I . e .
. g.
.. - . - __.... ~_______._--__-_e
~.
~
c . ;,. . . .
~
~- _
I . r% ,,.n .
- - .. ~
.' [', SQH-1
~
s '. . TABLE 15.3.1-2 . 4 LOOP SMALL BREAK A, ,' t - r;. : .. . Crse Anal'v sis Results - 2.0 Inch 3.0 Ineh 4.0_ Inch , f'
< ,4 -
1 Pea'$t clad temperature.(*F)
- 851.6' ' 2105.5- .' 1673.2 f* .
- 1
.,'. 11.50' ~12.0 11.75 i Peak clad location (ft) tsaximum local 2'/M 9.079' 1.074'- . .
r 20 react, ion (%). C.0348
. .v ; 11.75 ,
Local' 2r/M 20 location (ft) ,,
. g 11.5.',.t 6 ,,.
12.0 '.. ,. : 40.3 ' . <0.3 .,. <0.3 .i , I Total Zr/M 20 reaction (%), .,
'N/A-
.-e
~
~
N/A W/A 150t rod burst tim'e (seconds) t,
"'- N/A -, N/A- ,
Hot rod burst 1 kation W/A / .. . , . . .
. i t
e 1 Calculation Ass **=retionss . Core Power (WoG;Heatup Calculation)' (MWt) 102% of 3411 , l 14.075, t Peak Linear Power (kw/ft) 102% of , l 2.70 ,, ' i Peaking Factor (at License Rating) l 3 1050 Cold Leg Accumulator Nater volume (n'minal (ft )]~ ) . 615 Cold Leg Accumulator Pressure (psia) 8 15 l Steam Generator Tube' Plugging Invel (%) , j . f 4 6 , * + j . , i 4* .I ' .
~ '
j C , . t 9 ' a'.
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. 1 4 - 4 i
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- X -
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safety Injeetion Flowrate
- Figure 15.3.1-1
- r-
. Ys. Pressure .
j _- 4 i _~.
# er
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~
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- 9 *
'_._________--._____i__ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ - . , . . . ...._...,,,-.,_..m.,.,,,,, . _ , , , ,,,m_, - . _ _ , - . , , ,
l
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3 ~
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.. 0 . . .
l 0- - 100 - 200 300 400
~ '
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s.- , [
-.. . x
- - - ~' ~
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m,g v Figure 15.3.1-2 TVA 8.0 Inch Cold 143 Break Reactor Coolant System Depressurisation _ Transient. I
.s ,
l . J' # i
l i !
. j
'C. 2'488.
2249, x-3008. ]
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. age =
- 3. 488. ,988. 4200. 1600. 2008.
TIME (SCCI
+ 3 .
l [ .- - L 69* Figure 15.3.1-2 TVA 3.0 Inch Cold Leg Break i . Reactor Coolant System - l Depressurization Transient - y 023 ,
. - ~ - , - -~,-~---_---------__--_..u.---__,-,w -.--n.,-,-,.....----,,-,,-.-n,.-.,,,--,~m....,,--,,w,-.,.,a,e,v-,-,---m--
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.- r
- ~,
z .. j r
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{ { " 0.0 0 100 200 300 , 400
.- TIME (SECONOS) a.- .. .c ~
^9
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1 Figure 15.3.1-3 TTA 8.0 Inch cold Les Break Core .
*- Mixture Reight Transient
- r. .
4,
*~ %
l l.
. ___ -. x ,
. .. . i
. - . c o e ._ . , _ . , , . . , , . _ . , - , _ , _ , . , _ _ _ . .-. _ . . . . , , , , , - - , . . , _ . ,
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E TOP OF CO RE gy, -_ - . _ _ _ _ _ 3 - ' ! ,. g . . l . q . 3
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088. 1288. 1688. 2888. TIME' ISCCl
.. w -
as
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1 Fig 6re 15.3.1-3 TVA 3.0 Inch Cold Leg Break , Core Mixture Height Transient ' 1 .
=- - ;
025 , l l-
1 i
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( ... .
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* ?
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l
- I
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Figure 15.3.1-4 TVA 3.0 Inch Cold Leg Break
~
.- clad Temperature Transient
+ 027 -
e 4P u>w v,t-w- em,+ ee., ,- -,--e...,me- e - + ~ _ _ . _ _ _ _ - - -_ ,,e-. _ _.~e,'*wr
-- me----- -t
. , l
.r.
12.179 4 C_ , .
. fsPLAC6 1
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-g- ~
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4 4 4 ['
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? -300 --
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, . t l
~'
- .g. -406 . :::: ~
f L _
-s00 . .
_L 1 !- 0 100 - 200 300 .560 . s
..- ~
TIME (SECONOS)
~.
~
s'
. =
b" h s i
+
** +
~
' figure 15.3.1-5 TVA 8.0 Inch Cold Les Break Core Stasa' r w m.e. .
t.' ..
.. ~
d: ,. g I e . m +:r. - . . - , _ _ _ . , _ _ _ . bgD"
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an. - 4 l.1 .
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, er- .
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L
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- p. pg, ;
a> A_h --- o . - ,- .
- f. l 85 7 - 488.- BE8. ._ 8268. 1608, 2000.
I
. flME ISEC) .
I
~
2 . l 9
, Figure 15.3.1 5 TVA 3.0 Inch Cold Leg Break Core Steam Flow Rate
~
r.
- 029 s ,
, - - - - - . . - , - - , - . < . . , . - . - . . - , , , - - , , , - - - , -- .--,-.,..,...-.,--....--..~.---+-r-~--
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r 9 S-I dW8
't Figure 15.3.1-6 TVA 3.0 Inch cold I.eg Break
. Rod Film Coefficient
- 031 j, ,
ll k _ . _ , _ , , , _ , . , . _ . . . . _ , . . . . , , , , . . . , . ._ . . , , . , . , ,.. , , -yv y ,
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t 4 W6 a.. . 8% 9 Figure 15.3.1-J TVA 3.0 Inch cold Leg Break Not Spot Fluid Temperature .
~
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- 036.
e 9
i
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- 12.279 10 l
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.: s.*
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' 1 I I .
0 ' 0 500 1000 1500 2000 s
- TIE (SECONOS) .
~~
** . 1 4 f -
s '
- '1 Figure 15.3.1-10 TVA 4.0 ' Inch cold Les Break Reactor Coolant System Depressurization
., Transient U
~
, N .
x
, s.
s . .. .. - _ . _ . . . . , . . . _ . . . , _ , . . , - . . . . . - - . . - . _ _ , . . , . , ,,...-.1 ,.
. __ _ . . . ~ . . _ _ _ _ . _ _ . _ _ . . _ _ _ . - . _ - _ _ _ . _ _ .
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ISOS. , 1500. 2800. 2508, 5000. 5500. 4000. TIME t.stc) -
.- ~
.~
r
*w er*
Tigure 15.3.1-10 TVA 2.0 Inch Cold Leg Break Reactor. coolant System Depressurization Transient 03& -
% d e
4 4
. . . . . ~ . . . . - . , , - .- . . . . --.~.- _
^
w
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+
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t 1 g; '- g4 - , .: ,
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g . { . I000- - t.
-s.-
-~
500 - -
.u.
r I l I l n '
-0 250 500 . 750 1000 1250 1500
~
- Y ?- TIME (SECONOS) -
. e Figure 15.3.1-11 TVA 6.0 Inch cold Les Break Reactor Coolant System Depressurisation Transtant.
.D .
. e 4
4 4
=
44
,- . j
=. . 4 1 o' C ' 2488. 4 . 2388. -
. ~ -.
\
,, 2889. -
s
.E 1889. .
g - 1688 ' M l h1488. . n . k i , .
" 1200. ,
u _ 3 C [seea. 3
.x -
800.
~
. 2, .* -
. 600. s .
s - 400 -
- r
. D. 288. 400 600. 880s 10.00. 1200. 1400. 1600. 1800,
. , TIMc estes
' ~
+ .
~ . . -
-s . ..
~-
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, Figure 15.3.1-11 TVA 4.0 Inch Cold Leg Break Reactor coolant System
~
Depressurization Transient 000 e e
,.s
,x - ...
3 .; . ,
, , p -
.- e ji 7, i ,
~
W- , - -:
. /,: ..
c .. . q ,; , ,
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~,
3 ,
/
i _; . . l _ll ' y .
-'2500 . ',
.f . -
1 2000 - . g x
- g. -
= 00
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s w g j v . ~
- 1000 - !
.. . s ,
,s 5b0 s
) ,_ . .
. j
- l / e /
l,. .s o 0 100 2 300 - 400
, =
"W Es e
~
~
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- TIME, ECONOS) t . .
% e
'2 .
. .~,.. ..
-a 1:,, -
1; , l, . Figure 1 '.3.1-12 TVA 0.5ft Cold Leg Break Reac or Coolant
,'- System D ressurization Trans nt
[ j .X -
.. % J t .-
, ' i
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v.
~. ,
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-{(,
. . .~ ..
fe.Pt.AcA
/e .
N
. m. .. -;
ep- 4 20.0
. 3 s
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17.5 -
,. i
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g . .:
~
t [ 12.5 .
.f
. . W - ,,; . j .
3 i .- r
-t
. o" 10.,0 -
.~_.
7.5 .- - t t
,+
~
~. I I I s 5.0
~
0 ,500 1000 1500 2000
.. .u .
- flE ISECONDS) 1'- .
~
.~
s,
- g gsi/15u/
Figure 15.3.1 M m 4.0 Inch cold Les Break core 8F Mixture Haight Transient
% 4 N .
b y so 9 v
aW ' a a . C 51. i
~
L' 50.
, 1
- - r ..
. -29.
o c h 28. . g'* . . b 27. - y t6. - p y 25. . , _. Guus a z ( g - 1
.24. .
lld'
-25. .
a .. , l- TOP 3F CORE .
- 22. ~ .
~ ~~ % ==
~
- r I
' 20
_D. 600. ' 1880. 1500. 2000. 2500. 5000. 5500. 4000. .
~
. TIME (SEC) .
l
~
s . . -
.e, me .
- Figure 15.3.1-12 .TVA2.0IhchColdLegBreak --
Core Mixture Height Transient
,f I' 043 .
L
,m---
~
n 5 ' .r ( c- , F .
*- . I e . . ,,
12.279 14 t . y! . ', - {(.. V . fgycf - . l u . 15.0 ' V .
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~ --
- r.
2.5 - - r 2, .. ,
- r, p
l i.,l -l l s 0.0 '
- 0 250 500 750 1000 1250 1500
=
, TIE (SECONDS)
~
's . , ~ __
- _+
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. Figure 15.3.1 M- T7A 6.0 Inch Cold Leg Break Core T
g Mixture Height Transient k o 8 a.- e 8
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t.. x 52. . 58.
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= .
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. 998. 800. 1800. 1200. 1488. 1600. 1800.
TIME' (SEC) - me.
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Figure 15.3.'l-13 TVA 4.0 Intih Cold Leg Break , Core Mixture Heigh *, Transient
\
4 045 e 4
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' TIME (SECOMOS) ~
's .
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--1 I
. igure 15.3.1-1 TVA .Sft Col Leg Break Co e
. Mixture Heigh Transient ;
i e 1
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.,= .,e, 8 .-
8 g
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. 500 750 1000 . 1250 1500 1750
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TIME (SECONOS) t- ~~ '
. u .-
- - 9 #
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Figure'15.3.1 TVA 4.0' Inch Cold Leg Break Clad
.s 44 Temperature Transient
* '4
. g
- .. . _ _ - ~___-.--. - . ._. - . - - - . . __ . ___ .
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Figure 15.3.1-14 TVA 2.0 Inch'Coro Leg Break clad Temperature Transient - -
*% r 4
048 : e f*" q 4
.g s
. . . - . - . . . - . - . - = . . .
J
- 1 1
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=
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SN . - 7M. . SN. M. . lM8. 81N. ING. 1500. Idea. 1800. 14N . 1799. I GN . _... , TIME ISCCI r
, = '
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Figure 15.3.1-15 TVA 4.0 Inch Cold Leg Break
, e Clad Temperature Transient l
. P*
. Yv
.' 6.
1
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~.< _.
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.5ft!CojdLegBeak Figu e 15.3. 8 TVA - j'
/ .
C'.dTenne:puttTr.sient
u< , .
.- SQN e
- 115.4 CONDITION I LIMITING FAULTS ,
Condition IV occurrences are faults which are not expected to take place, but are postulatef because their consequences would include the potential for the release of significant amounts of radioactive material. These _ are the most drastic.which must be designed against and thus represent limiting design cases. Condition IV faults are not to cause a fission . . product release to the environment resulting in an. undue risk to public .
' ~
health and safety. in excess of guideline values of 10 CFR Part 100. - A
- single Condition IV fault is not to cause a consequential loss of 1
, required functions of. systems needed to cope with the fault including j those of the Emergency Core Cooling System (ECCS) and the containment. 1 For the purposes of this report the following faults have been classified :
i in this category: ,
- 1. Major rupture of pipes containing reactor coolant up to and L jncitfOlng double ended rupture of the largest pipe in the Reactor e l
lL
' Coolant System (loss of coolant accident), l
- 2. Major secondary system pipe ruptures. l 3." Steam generat'or tube rupture.
I 4. ' Single reactor coolaht pump locked rotor.
- 5. Fuel handling accident. .
,Cp
- 6. Rupture of a controlrod mechanism housing (rod cluster control - .
assembly ejection).-- . l- .
- The analysis of thyrold-and whole body doses, resulting from events leadirig to fisston product release, appears later in the Safety Analysis
-Report. The fission producl inventories which form a basis for these calculations are presented in Chapter 11 and Section 15.1. The Safety .
'.-: ; Analysis Report also inclinde's the~ discussion of systems interdependency contributing to limiting fission product leakages from the containment
. .following a, Condition 4V occurrence. .
15.4.1 'M'aior Reactor Coolant System PIDe Ruotures (Loss of Coolant I Accident) ,
, ggmg A3
?
. 3w
! TheAnalysis'specifle7by10'CFR50.46"6ccep,tanceCriteriaforEmergency - Core Cooling Systems for Light Water Nuclear Power Reactors", is presen :ed in this section. The results of the loss of coolant accident analys#s W e shown in Tabl ~15.4.1-1 c d ";.4. f and show5 compliance ! with tee-Acceptance Criter a. The description of the various aspects of
. ~
;;rn ed U"'. % nui G -
the ,LOCA analysis is given in 1teference534:n".: ,,deted 6 im.ius
=~ie ! ( r ~ t ea u ucap on7a, "a d : M 2.f ...
7
- . , . . . .. .$. . [.U ' . _b . N...NI. _, __'_ . h_k_* * " T. , dli. $_ f. . 'b"2_ h. ' r . . .
, f
, 4 8, M o Y 1 L
0117F/COC4
- O
y, TX ,, j g < r ,
.. _ 1 SQN-6
- ~
e .
- The boundary conside" red for loss of ' coolant accidents-is the Reactor-g^. '
- Coolant System (RCSL or any line connected to the system up to the first ,
= ,7 i l
C closed valve.. .= . l Should a major break occur, depressurization of the Reactor CDolant - Sy' stem'results in a pressure decrease in the pressurizer. A reactor trip , signal occurs when the pressurizer low pressure trip setpoint is -
< r- reached. A Safety Injection System signal is actuated when the .
'.. appropriate setpoint is reached. These countermeasures will limit the -
consequences of the accident in two ways: s -
.1~. ' Reactor ' trip and borated water injection complement vold formation
. in ~ causing rapid reduction of power to a residual . level ;
corresponding'to fission product decay heat. ,
- 2. - Injectkn' of borated water provides heat transfer from core and
- prevents excessive clad temperature.
At th's beginning of the blowdown phase, the entire RCS contains subcooled liquid which transfers heat from the core'by forced. convection with some I fully _ developed nucleate bolling. After the break develops, the time to p departure from nucleate. boiling is calculated in a manner consistent with = ' ' JKp~ pend 1x.K of 10.CFR 50 (1).. Thereafter..the heat transfer is based on L
-local conditions with' transition boiling and forced convection-to-steam as the major.,he'at transfe~r mechanism.
l-l- g hen he R pres re fal below e uppe head-1 jection ressur , the g- V
~ lg6 acc lat beg 1 to in et bora d wate . direct y into e reac or ,
" 7' i
(~ 'upp r he regt . . Thi water 1 direc d from he uppe head d r.ect.1 t all b t+eig pert p. al ass blies n the re via e RCC utde
-t bes a d UNI' upport' olumns. This f ow pro des add tional ore oo11 during the b down p ase of he tran lent.
WhenIthe:RCS pressure falls below the-accumulator tank pressure, the cold - b leg accumulators begin to inject borated water. The conservative .,,
, assumption is ma.d.e that inj'ected accumulator water b) passes the core and goes out through the break until the termination of bypass. This conservat1[a is again-consistent with Appendix K of 10 CFR 50.
15.4.171'. Thermal Analysis , !- 15.4.1.1 1 Teestincheuse 'Perforrance Criteria for Emeroency Core Coolina Systes , .
'Tne reactor is designed to withstand thermal effef:ts caused ~by a loss of
~
~ coola.nt ac'cident including the double ended severance of the largest
- Reactor Coolant System pipe. 'The reactor core and internals, together I ,
withethe ECCS are designed so'that the'rtactor can be shutdown safely.and - the efsential. heat transfer geometry .ot the ctre can te preserved following the accident. - f 15.4-2 0117F/COC4 O O
, ;s - M . ,<., . g , ' * . de, a. . ,
~
t i i c
-l < -
The ECCS even when' operating during the injection mode with the most _"_.'. l
' C"
-severe single active failure, is designed to meet the Acceptance ,
. . Criteria-(1).- ' -
- v" 15,.4.1.1.2 Method of Thermal Analysis g
.The description of the various aspects of the LOCA analysis is given .
,4@ .NCAP-8339.[33). This document describes the majorahenomena modeled, the y
interfaces among the computer codes and features of the codes which s l M~, s maintain compliance with the Acceptance Criteria, Differences.between ' the approved Non-UNI Westinghouse Appendix K Model and the model used for The thermal. , these analyses.are reported in WCAP-8479 Revision 2. analyses reported in t ' temperature based'on T..... The' UNI accumulatorLpressure set point l sures UNI ~ectuation prior to upper head fluid flashing. 15.4.1 .1 1.3' Containment Analysis , 1 L 'The containment _ pressure analysis is performed with the LOTIC-2 [373 ' h code. The transient pressure computed by the LOTIC code can be entered g h5N(Tn the'fiTLOOcode for the purpose of computina' the reflood transient. presented in Figure 15.4.1-1. 4 W C^ '; . ;; a ;. ~ -" n-p The - containment data used in'the containment pressure analysis to determine the ECCS backpressure We preseHted in Tables 15.4.1-4 and 15.4.1-5 The mass and energy release rates used for the containment backpressure
- [U
. calculation as.a function of time during blowdown, are given in -
I I
-L Table 15.4.1-6: -
15.4.l:14 Results o large Break Spectrum
'~
The:following four breaks were Investigated in this analysis: [,- Double-ended cold leg bieak (Co 0.8),-imp mix' Double-endeif cold leg break (Co 0.8), per mix '
. Doubfe-ended cold leg break (Co = 0.4), imp mix Double'-ended cold leg break (Co 0.6),~ imp mix .
l~ Figure 15.41-2 depicts the SATAN control volume scheme and can be used as a referenie for Re tfansient resultsi The transient results are g presented in. Figures 15.4.1-3 through'15,4.1-57 for the four breaks as l_isted'previously. The following nonmenclature applies: -
. per silx: perfect mixing in upper head durin'g UNI. water injection
, . imp mix: imperfect mixing in upper head during UHI water inject 1on
~
2 - flowrate in el 1, 2: flowrateat..towerha[fandmidplaneof core, respectively - 2 uce khre hcer T
- m. 4.
15.4-3 0117F/COC4
. r- _
W. .?G - - -
~ '
(. . n . SQN-4
' Z! - flowrat n el 3, 4: flowrate t upper half and to of core, 1 ,C respec ely s =/
lo. pha, v' d in el 1,.2: v id fr ction in 1 er half of core 3-foot
~
sections, respectively ., I, '
- lo.' pha void in el 3, 4: void fraction in upper h if of core -foot secti , respectively -
/
~
ovrate in el 43 '44: lowrate fr cold g and UHI ' Z- -
-- . ccumulators, re pectiv ly The time sequence'of eve s fcr the analys s descri ed below a shown in Tables 1 .4.1-7 and 15. 1-8; ~ ables 15. 1-1 and 15.4.1-2 p sent t I peak c1 dding temperat es an hot, spot tal rea tion for a pectru of-large tak sizes.
K The S AffVI analys1 of th loss of oolant ac ident is rformed t 102 p reent of the ore lip'ensed pow r. The cdre therma transien 'is - also erformed at is power level. . The peak /' linear po r, the p aking fact r of the lic se application ower lev , and cor power used in the .'" goal ses are giv in Table 15.4. -1 and 15 .1-2. S ce there s. margin {- bet een the val .of the peak li ear power ensity u d in this analysis L I ' L an the value pected/1!ioperaion,alowerpeakc d tempera ure will be obtained.,by using he peak .inear pow f density xpected d ring o eration. he plant'p rameter used i this anal ~ isare1ste$ int le 15.4 -9. , )-- he net of ect of hese-in ut parameters is co ervative f r the L l ' 'J , analysis, r the ollowtn rtAsons:
. 1. In ial .sys em con it'lons are hosen s as to maxi ize stor d energy
?l the RC 'and fu .,
-2.- old le /accumu tor parameters are elected to cover t potenti =.
impact f ()) nim 0m ECCS /bellvery rate resul ng in extende i period f ste / water a P ) maxi m ECC 4
- delivary, rate leading to /uring or ecumula r flood emptyinand. -prior bottom core recover .
l 3. UNIj'vol.ume are select d to ma mize the u per he reheat t me for . the~perfe t. mixing cas,e.snd a imize the I syst m's contr bution
.~ to' core ooling for he imper ect mixing ase. -
'For the.res ts discusse below the hot s is.de'ined to b the
. -locat' ion o maximum peak claddi g temperatu e. Th s locatto is given in
.1-1 and 15(4.1-2 f r each brea size nalyzed.
Tables 15 .
/ .- -
ya 6 e 15.4-4 0117F/COC4
+. - .-
p _ 4 d I v . j b <m -SQN-6' - E l 4 Figures 15.4.1-3 through 15.4 1-57 present'the trapstents for the ' 1 0' principal par er,s for the break sizes analyzed / The followin items are noted: -
-Fiiures'15.'4.-3-Thefoiowingquantitisare resented at t claddi g g through 15.4 1-14 burst Jocation and at he spot (locati of max um ich are on th hottest uel .
clad temperature) bo of rod (hot rod):
. l
/ u
~
i
-. 1. ./ fluid 2./ mass v. quality
/ I elocit /-
3.. heat transf coefficient. The heat trans er coefficient show is calcul ted by ' t the LOCTA IV odel ,/ . . L~ F gures 15:4.1-15 The system ressure shown is th calculated' pressure ' hrough' 15.4.1-34 in the cor . Core flowrates a core vold/ fraction ar also pres ted. ; 8 E Figures 15.4.1-35 These fl ures show the hot e ot ' cladding' temperatur - through 15.4.1-41.~.transie t and the cladding /:emperature transient at the burst ocation. The flui temperature /shown is al o for t e hot spot and buts nota .on of tne figures s location. defined tr)jThe
. Tablenodal 15.4
-10.
Figures 15.4.1-42 Th' e fleures show the core reflood/ transient. ' l [ C'/) '.through 15.4.1-49
,i j
Figures 15.4.1-50 hese figuris show e Emergency Core Cooli System i- " through 15.4.1-57 flows for.all case analyzed. Both UHI an cold leg accumulators flow rates are included in t figures. ' I ~ . l
~~
As described ear ter the cold . leg accumul tor deliver g l during blowdown is discarded ,until the e d of bypass s J t calculated T e cold leg accumulator f ow assumed the sum of th injected in the intact cold legs. , , 1 ; I . ~ The centairt ent pressure tran ent for one of the case analyzed (C = l l: 0.8, imper ect mixing) is pr ented in Figure 15.4.1 . Figures 1 .4.1-67 Il l , and 15.4. -62 show the heat emoval. rates of the upp r and lower 'c part- g ment hea strTks respective : Jhe heat transfer mod I used'is des 'ribed
-in Refer nee'37. Figlire .4'.1-63 gives,the flowr te at the exit of the j ice co enser drains. F ure 15.4.1-64 prese'nts etemperature/ -
i transt nts in both the per and lower compartme ts. Figures 15.4.1-6 l f 4nd 1 .4'.l'66 lllustra the. heat removal rates jthesumpandthe.1 I cond nser drain. Tot lower compartment heat emoval is the summati n - of ,e, rates'given i Figures 15.4.1. 62, 15.4 -65 and 15.4.1-66. [ ~
"~'
The results of the break spectrum shoNhat he d'ouble-ended, cold eg i gulliotine break with a discharge coeffic nt of 0.6 and Imperfe t ! - mixing in the u er head, .is the worst bre k in terms of calcula ed peak eladding - temperiiture.
~
I ' 15.4-5 Oll?F/COC4 O O
.,g", of ,% ' e
y , j
- . *** INSERT A *** .
15.4.1.1.2 Method of Thermal Analysis , Descriptions of the various aspiscts of the LOCA analysis are provided in References-2 and 49. These documents describe the major phenomena modeled, the interfaces among the computer codes, and the features _of
'the codes which serve to maintain compliance with the acceptance
-criteria of 10CFR50.46. - - s -
[ 5he analysis of a'large break.IhCA transient is divided into three I phases:' Blowdown, Refill,-and Reflood. A series of computer codes has L Dean developed to analyse the transient based on the specific phenomena which govern each phase. During the blowdown portion, the SATAN-VI code
-(Reference 35) -is used to' calculate the RCS pressure, enthalpy, density, ~
and mass.and energy flows in the primary system, as well as the heat
- transfer between the.. primary and secondary' system. At the end of the
- -blowdown, information on the state of the system is transferred to the l
WREFLOOD code (Reference 36) which performs the calculation of the
' refill period to bottom of core (BOC) recovery time. Once the vessel has
, refilled to the bottom of the core, the reflood portion of the transier.t .
, begins'.' The BASH code (Reference 2) is used to calculate the thermal-hydraulic simulat, ion of the RCS for the reflood phase. Information con'cerning The cori boundary conditions is taken from all of}}