ML20248J755
| ML20248J755 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 04/06/1989 |
| From: | William Cahill TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TXX-89172, NUDOCS 8904170039 | |
| Download: ML20248J755 (37) | |
Text
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l Log # TXX-89172
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File # 10010 J
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7t/ ELECTRIC 915.2 (clo) j f
Nat,?v?a rra,isant
%d16, GM U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 INFORMATION TO SUPPORT TECHNICAL SPECIFICATION CHANGE FOR LOW-LOW STEAM GENERATOR WATER LEVEL i
Gentlemen:
i During the Technical Specification (T. S.) review meetings that took place l
from February 13 through 24, 1989, TV Electric agreed to submit a safety analysis which demonstrates the acceptability of a low-low steam generator water level safety analysis limit of 0% of the narrow range instrument span.
I This analysis is used in the development of the setpoints identified as i
Functional Unit 6.b.in T. S. Table 3.3-3, sheet 4 and Functional Unit 13 in i
T. S. Table 2.2-1.
In accordance with that agreement, enclosed is an advance description of a FSAR change which will be included in the upcoming FSAR Amendment 76.
If you have any questions on this material, please do not hesitate to contact-me or my staff.
l Sincerely,
=
William J. Cahill, Jr.
RLA/vid Enclosure c - Mr. R. D. Martin, Region IV Resident inspectors, CPSES (3) pod
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400 North Olive Street LB81 Dallas, Texas 73201
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J COMSISTS OR THf FOL.LOLUfUG FSAR l tents :
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Ta ble.
15,o y (g; 2)
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s CPSES/FSAR plant thermal kinetics, RCS including the natural circulation, 57 pressurizer, steam generators and feedwater system. The digital i
program computes pertinent variables including the steam generator level, pressurizer water level,.and reactor coolant average temperature.
The assumptions used in the analysis are as follows:
~
1.
The plant is initially operating at 102 percent of the engineered safety features (ESF) design rating.
2.
A conservative core residual heat generation based upon long term operation at the initial power level preceding the trip.
3.
A heat transfer coefficient in the steam generator associated with RCS natural circulation.
4.
Reactor trip occurs on steam generator low-low level.
No credit 57 C
MM4 DM'n d is taken for immediate release of the control rod drive mechanisms caused by a loss of offsite power.
MW 0 'W P3%,
a y e,f) 5.
Euxiliary feedwater is delivered to two steam generators.
73 3
6.
Auxiliary feedwater is delivered by either:th matcr driverr 73
^uxlliary feedwater pump or the turbine-driven auxiliary fepwaterp 7.
Secondary system steam relief is achieved through the steam 57 generator safety valves.
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C 15.2-17 Amendment 73 August 5, 1988
)
- s CPSES/FSAR 73
- 8.
The initial reactor coolant average temperature is 6.50F higher than the nominal ESF value, and initial pressurizer pressure is 30 psi higher than nominal.
73 9.
The lower auxiliary feedwater flow rate results in a larger amount of coolant expansion into the pressurizer.
The pressurizer power operated relief valves are assumed to function I
normally to maintain the peak reactor coolant system pressure at or below the actuation setpoint (2350 psia) throughout the Q ransient.
Plant characteristics and initial conditions are further discussed in Section 15.0.3.
i l
15.2.6.2.2 Results 57 The transient response of the RCS following a loss of AC power is shown in Figures 15.2-9 and 15.2-10.
The calculated sequence of
. events for this event is listed in Table 15.2-1.
74 The first few seconds of the transient following receipt of a reactor trip signal will closely resemble a simulation of the complete loss of flow incident (see Section 15.3.2), i.e., core damage due to rapidly increasing core temperatures is prevented by promptly tripping the reactor. After the reactor trip, stored and residual decay heat must be removed to prevent damage to either the RCS or the core.
1 57 The LOFTRAN Code (3) results show that the natural circulation flow l
available is sufficient to provide adequate core decay heat removal following reactor trip and RCP coastdown.
i l
Amendment 74 15.2-18 October 14, 1988
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- q CPSES/FSAR i
r 15.2.7.2 Analysis of Effects and Consequences l
15.2.7.2.1 Method of Analysis i
A detailed analysis using.the LOFTRAN Code [3] is performed in order to obtain the plant transient following a loss of normal feedwater.
The simulation describes the plant thermal kinetics, RCS including
-)
j, natural circulation, pressurizer, steam generators and feedwater
.j system.
The digital program computes pertinent variables including the steam generator level, pressurizer water level, and reactor j
coolant average temperature.
Assumptions made in the analysis are:
1.
The plant is initially operating at 102 percent of the enginee' red safety features (ESF) design rating.
I 2.
A conservative core residual heat generation based upon long term
(
operation at the initial power level preceding the trip.
3.
Reactor trip occurs on steam generator low-low level.
I I
4.
The worst single failure in the Auxiliary Feedwater System 57 occurs.
l 5.
^exiliary feedwater is delivered te four steam generators.
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Quxiha feedaa ter. is debt.cres{
3, 860 fo hur Steam Jen era lo rs G}9iks{ a Ctean, /ine ba6N rescue of M % fsia..
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15.2-21 August 5, 1988
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'CPSES/FSAR-
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'An additional assumption'made for;the-loss of normal feedwater 57 evaluation is that only the pressurizer safety valves are assumed to function normally.
Operation of the-valves maintains peak RCS pressure at or below the actuation setpoint (2500 pounds per square
. inch absolute;(psia)) throughout the transient.
.Since.the two Comanche Peak units will have different steam generators 5
(see Section 5.4.2), the effect of this' difference has been considered in the analysis. Both types of steam generators are integral preheater models.
The major difference, from the standpoint of accident ~ analysis for this event, is the slightly higher secondary
' side mass as a function of power for the 05 (Unit 2) model.
In order 1
to maximize the time until reactor trip on low-low steam generator
{
1evel occurs and to insure that the analysis is valid for both units, j
the initial steam generator secondary mass was assumed to be 110's of thehigher05 mass.JThelow-lowsteamgeneratorwaterleveltrip
~
setpoint was assumed to be the same mass (ib. mass) for both units i
(see Table 15.0-4 Note that while a higher secondary mass (larger heat sink) is, in general, a benefit for primary side heatup eW]
the assumption of a higher initial mass results in a delay of the trip signal, and thus produces a more severe transient.
In addition, all steam generators for both units will be equipped with 5
i separate feedwater connections for injection of auxiliary feedwater t
and main feedwater at low power operation.
The major effect of injecting auxiliary feedwater into the upper section of the downcomer is that most of the flow will bypass the preheat. region due to the higher resistance to flow in.the preheater.
This will result in a slight decrease in heat removal capability.
However, the auxiliary feedwater injection point is now much closer to the steam generator, resulting in a much smaller volume of hot feedwater which must be purged before the colder auxiliary feed enters the units.
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-- Er2-23 August-5-1988
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CPSES/FSAR Plant characteristics and. initial conditions are further discussed in 3
Section'15;0.3.
Plant' systems'and equipment which are available to 7
mitigate the' effects of.a loss of normal feedwater accident are discussed in Section;15.0.8 and listed'in Table 15.0-6.
Normal
~
j reactor control. systems are' not required to function.
The Reactor Protection'-System is required to~ function following'a loss of normal
..feedwater as analyzed here.
The Auxil'iary Feedwater System is
'57 required-to deliver a minimum auxiliary feedwater flow rate.
The auxiliary feedwater flow rate assumed for the Loss of Normal Feedwater g40
- analysisis'40@. gal /mi[No'singleactivefailurewilTprevent N
operation of_ any p stem,Lrequired.to function. A discussion of ATWT.
considerations is presented in Reference [2].
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Results f gg pca, Figures 15.2-11 and 15.2.12 showlthe significant plant-parameter transients following a loss of normal feedwater.
Following the reactor and turbine trip from full load, the water level' in the steam generators will fall due to the reduction of steam generator void fraction and because steam flow through the safety 57
-valves continues to dissipate'the. stored and' generated heat. One minute following the initiation of the low-low level trip, two motor-driven auxiliary feedwater pumps or one turbine-driven auxiliary feedwater pump is automatically started, reducing lthe rate of water level decrease.
1 73 The auxiliary feedwater flow rate for this event is higher than that
,-for the loss of nonemergency AC power event (section 15.2.6) due to the additional heat input to the coolant from the reactor coolant pumps.
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I Arendment 73 15.2-24
. August 5, 1988 l
~
9 CPSES/FSAR ^
8.
A conservative feedline break discharge quality is assumed prior to the time the reactor trip occurs, thereby maximizing the time the trip setpoint is reached. After the trip occurs, a saturated
-liquid discharge is assumed until all the water inventory is discharged from the affected steam generator.
This minimizes the heat removal capability of the affected steam generator.
57 9.
Reactor trip occurs on steam generator low-low level.
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The Auxiliary Feedwater System is actuate y
e low-low steam j
73 generator ~wat~er~1evel signal.
The Auxiliary Feedwater System is j
assumedtosupplyatotalof430gallonsperminute(gpm)[et'een w
two unaffected steam generato m a motor-driven pump. M wuc.
% :t,hurbine-driven pump,i able of supplying 430 gpm to three intact steam generators, "at" then te two intact-st+am
/gencceim..,.) A 60 second delay ssumed following the low-
/ h ceA2.ideu2h ow level signal to allow time for startup of the emergency U
"b diesel generators and the auxiliary feedwater pumps.
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ceyfrad & furge) ee 73 Approximately 106 secondsp= =:umed before the feedwater lines
- rg
- d and th relatively cold (1200F) auxiliary feedwate re ccui enteraff the unaffected steam generators.
4 73 11.
Thirty minutes after the reactor trip, an additional 265 gpm is assumed to be supplied to the third intact steam generator by operator action.
12.
No credit is taken for heat energy deposited in RCS metal during the RCS heatup.
77ds cz.:ss &
//rt cauxc >rMtt MSN anyas as<dts % doelfadny a OtY Nona gemeb-et Amendment 73 15.2-30 August 5, 1988 I
CPSES/FSAR RCS pressure will be maintained at the safety valve setpoint until 57 C
safety injection flow is terminated by the operator or until AFW flow is increased to the intact steam generators as mentioned in Section 15.2.8.2.
The reactor core remains covered with water throughout the 71 transient,endgaterrelief*duetothermalexpansi, s p;^= = t b y (
.f the heat removal capability of the Auxiliary _Feedwa er System 3y rea.olk &dnus& 6 ytvo<& ly L Sala 1 ly'eekw QLQ Q
The major difference between the two cases analyzed can be seen in the plots of hot and cold leg temperatures, Figures 15.2-16 through 15.2-18 (with offsite power available) and Figures 15.2-23 through 15.2-25 (without offsite power).
It is apparent from the initial portion of the transient (<300 seconds) that the case without offsite power results in higher temperatures in the hot leg.
For longer times, however, the case with offsite power results in a more severe rise in temperature until the coolant pumps are turned off and the Auxiliary Feedwater System is realigned.
The pressurizer fills more rapidly for the case with power due to increased coolant expansion resultingfromthepumpheataddition$gth
,hr.u, m water is relieved j
foyith= case d eviously stated, the core re ins covered with 57 water for both cases.
w,1%,o((ggfe % czy m /g /,/q f g 15.2.8.3 Conclusions Results of the analyses show that for the postulated feedwater line rupture, the assumed Auxiliary Feedwater System capacity is adequate to remove decay heat, to prevent overpressurizing the RCS, and to prevent uncovering the reactor core.
I 15.2.8.4 Analysis of Radiological Effects and Consequences Radioactivity doses from the postulated feedwater line rupture are less than those previously presented for the postulated steam line break.
All applicable acceptance criteria are therefore met.
57 L
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15.2-35 August 5, 1988
_____--_-__-___-___-----_-__A
CPSES/FSAR~
TABLE 15.0-4 (Sheet 2)'
TRIP POINTS AND TIME' DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES Limiting Trip Trip Point Assumed Time Delays function In Analysis (Seconds)
Low reactor coolant flow 87% loop flow 1.0 (from loop flow detectors)
..~
Undervoltage trip 68% nominal 1.5 Turbine trip Not applicable 2.0 Low-low steam generator 34.6%*-(Unit 1) 49 j
level and 0% (Unit 2) 49 of na7 row range 49 level span 49
~
49 High steam generator 90% (Unit 1) and 2.0 49 a
level trip of the 81% (Unit 2) of 49 feedwater pumps and narrow range level 49 closure of feedwater span 49 systems valves, and 49 turbine trip 49 The basis for the Unit 1 limiting setpoint is the loss of Normal Feedwater analysis.
The setpoint used in the Feedline Break 73 analysis was assumed to be 15%
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L-j Amendment 73 l
August 5, 1988
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'CPSES/FSAR
' TABLE 15.2-1 (Sheet 5 ef 9) 1 TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 1
Time Accident Event (seconds)
Loss of non-emergency Main feedwater flow 10.0
-l AC power stops q
Low ste'm generator I"7'8 a
water level trip Rods begin to drop W
6 f.O Reactor coolant pumps SG2E. 7I. 6 begin to coastdown Peak water level in S E F: 72.0 73 pressurizer occurs
%d 4*e steam generatorsg 44M4127.8 73 begin to receive g gpm 73 from auxiliary feedwater system 73 Core decay heat N
a335 73 decreases to auxiliary feedwater heat removal capacity a
i Amendment 73 August 5, 1988 J
CPSES/FSAR TABLE 15.2-1
-l (Sheet 6of9)
TIME SEQUENCE'0F EVENTS FOR INCIDENTS WHICH CAUSE A DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Time Accident Event (seconds) l Loss of normal feedwater Main feedwater flow 10.0 flow-stops l
l Low steam generator FfM 47.8 l
water level trip l
Rods begin to drop 4#;* 6 7.8
[
Peak water level in it @ 72.0 pressurizer occurs Four steam generatorg 1493 ##'O 73 begin to-receive y gpm from auxiliary feedwater system h ee Note k2 a 3 2. 0 Core decay heat N
decreases to auxiliary feedwater heat removal capacity j
i I
l Amendment 73
{
August 5, 1988 l
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CPSES/FSAR TABLE.15.2-1,
-l (Sheet 7 of 9)
TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM l
' Time Accident Event (seconds)
Feedwater system pipe break
- 1. With offsite power Main feedline rupture-10 available occurs Low-low steam generator 3M 31,'A level reactor trip setpoint reached'in ruptured steam' generator i
Rods begin to drop ikht 36.S Pressurizer safety valve 98 89,5 setpoint reached Steam generator safety 96 39.5 valve setpoint reached in intact steam generators Auxiliary feedwater is 9297 73 delivered to two intact 73 steam generators 550,7 Low steam lire pressure 468 6 l73 setpoint reached in ruptured steam generator i
567.7 All main steam line 3;t379 73 isolation valves close Peessa reser-ander 1 ;w relle f begi ns Amendment 73 August 5, 1988 i
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1 CPSES/FSAR TABLE 15.2-1 (Sheet 8 of 9)
TIME SE0VENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 1^
l Time Accident Event (seconds) l Core decay heat plus
-M BM-l pump heat decreases to
~ IM auxiliary feedwater heat removal capacity
- 2. Without offsite power Main feedline rupture 10 occurs Low-low steam generator M 3 p/2, level reactor trip setpoint reached in ruptured steam generator Rods begin to drop, power 3tt N,1 i
lost to the reactor coolant pumps Pressurizer safety valve 98-895 setpoint reached i
eam generator safety 9B 2 9. O 2
valve setpoint reached in (intactsteamgenerators Auxiliary feedwater is 4 8 d 9 6 2-73 delivered to two intact 73 steam generators i
1 Amendment 73 August 5, 1988 j
CPSES/FSAR TABLE 15.2-1 (Sheet 9 of 9)
TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A DECREASE l
IN HEAT REMOVAL BY THE SECONDARY SYSTEM Time Accident Event (seconds) i Low steam line pressure W V17 73 l
setpoint reached in ruptured steam generator 8 s1 All main steam line 44&S-Y2 9 73 isolation valves close Core decay heat plus
~1850 pump heat decreases to auxiliary feedwater heat removal capacity i
Note 1:
DNBR does not decrease below its initial value.
l i
Note 2: Analyses assume 600 gpm for conservatism during accident conditions.
Four steam generators would receive more flow from the Auxiliary Feedwater System.
Amendment 73 August 5, 1988
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FINAL SAFETY ANALYSIS REPORT UNITS 1 AND 2 i
Pressurizer Pressure and Water Volume Transient for Loss of Offsite Power Amendment 73 FIGURE 15.2-9 August 5, 1988 052 A 258M t
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FINAL SAFETY ANALYSIS REPORT UNITS 1 AND 2 Steam Generator Mass and Reactor Coolant Loop Temperature for Loss of Offsite Power Amendment 73 FIGURE 15.2-10 August 5, 1988 05Jo25890 2
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FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 PRESSURIZER PRESS. & WATER VOL.
TRANS FOR LOSS OF NORML FEED.
AMEN 0 MENT 57 DECEMBER 20, 1985 i M RE 15.2-11
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FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 STEA.M GENERATOR MA55 AND REACTOR COOLANT LOOD TEMPERATURE FOR LOS$ OF NORMAL FEE 0 WATER AMENDMENT 57 OECEMBER 20, 1985 MGURE 15.2-12
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FINAL SAFETY ANALYSIS REPORT UNITS 1 AND 2 Main Feedline Rupture With Offsite Power Available f
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August 5, 1988 052 A 25890 3
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FINAL SAFETY ANALYSIS REPORT UNITS 1 AND 2 Main Feedline Rupture With Offsite Power Available Amendment 73 FIGURE 15.2-14 August 5, 1988 052 a 258904
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FINAL SAFETY ANALYSIS REPORT UNITS 1 AND 2 Main Feedline Rupture With Offsite Power Available Amendment 73 August 5, 1988 FIGURE 15.2-16 012 A 25490 6
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FINAL SAFETY ANALYSIS REPORT UNITS 1 AND 2 Main Feedline Rupture With Offsite Power Available l
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FINAL SAFETY ANALYSIS REPORT l
UNITS 1 AND 2 7
Main Feedline Rupture With Offsite l
Power Available Amendment 73 August 5, 1988 FIGURE 15.2-18 01).A.25890 8
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FINAL SAFETY ANALYSIS REPORT UNITS 1 AND 2 l
Main Feedline Rupture With Offsite Power Available Amendment 73 FIGURE 15.2-19 August 5, 1988 051 A 25890-9
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052.A 25690-17 I
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FINAL SAFETY ANALYSIS REPORT UNITS 1 AND 2 Main Feedline Rupture Without Offsite Power Amendment 73 August 5, 1988 FIGURE 15.2-20 092 A 2589010
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COMANCHE PEAK S.E.S.
FINAL SAFETY ANALYSIS REPORT UNITS 1 AND 2 Main Feedline Rupture Without Offsite Power Amendment 73 FIGURE 15.2-21 August 5, 1988 D12 A 2549011
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FINAL SAFETY ANALYSIS REPORT UNITS 1 AND 2 Main Feedline Rupture Without Offsite Power Amendment 73 FIGURE 15.2-22 August 5, 1988 012 A 23890 62
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FINAL SAFETY ANALYSIS REPORT UNITS 1 AND 2 1
Main Feedline Rupture Without Offsite Power l
Amendment 73 August 5, 1988 FIGURE 15,2-23 012 A 2589013
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COMANCHE PEAK S.E.S.
FINAL SAFETY ANALYSIS REPORT UNITS 1 AND 2 Main Feedline Rupture Without Offsite Power Amendment 73 FIGURE 15.2-24 August 5, 1988 012-A 2589014
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Main Feedline Rupture Without Offsite Power Amendment 73 August 5, 1988 FIGURE 15.2-25 1
052 A 2189011
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FINAL SAFETY ANALYSIS REPORT UNITS 1 AND 2 Main Feedline Rupture Without Offsite Power Amendment 73 FIGURE 15.2-26 August 5, 1988 052 A 258% 16
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052.A 2589014 b
CPSES/FSAR Q212.75
-For the reactor coolant pump locked rotor and shaft break-
{
events and for the feedwater line break event, does the analysis assume water relief from the pressurizer safety valves? If so, provide justification for the: water relief rate assumed in.the analysis and state if the basis for the -
hydraulic loads used to analyze the mechanical design of.
I
.the valve, discharging piping, and their supports include i
water relief loads.
R212.75 No water relief through the pressurizer safety valves I
occurs for the reactor coolant pump locked rotor and shaft break events. For the feedline break events, wate g ef dr.es occur. @s shown-on figures 15.2-12 and -13,)pe naximum water relief rate through the three.(3) safety j
valves is a total of approximately 1 ft /sec. The flowrate' 3
is calculated based on the homogeneous equilibrium model for saturated fluid, which gives the most conservative This model indicates a' maximum available relief rates.
3 relief capacity of 17.4 ft /sec. at 2500 psia, much greater.
i than that predicted in the analysis of.the feedline break.
event.
The basis for analyzing the mechanical design of the j
Reactor Coolant Pressure Boundary Class 1 Piping and
~~
supports are discussed in Section 3.9N.1.
Class 1 valves, q-
-- sucLas the-pressurizer safety. valves,_are. d.iscussecLin_._
_ ___ l Section 3.9N.1.4.5 and 5.4.13.
The pressurizer safety
{
valve discharge piping and relief tank are non-nuclear safety.
d N
]
The pressurizer safety valves at Comanche Peak have been analyzed for the MFLB conditions consistent with the work performed for WCAP-11677. The conclusion in V', AP-116'17 that the Crosby SM6 valves can pass slightly subcooled water as-e-
_ 9 1-"- '; to three times without damage applies to the Comanche Peak valves.
p at le esj s
212-139 MAY 31, 1979 -
,eA.-
CPSES/FSAR'
'{,
Westinghouse setpoint metholodolgy) within 5%
of the top or bottom of the instrument range 49 will respond any differently than any other protection function.
Because large steam generator pressure changes are not expected before trip, only the reference leg heatup effects need be considered, and not the effects of system pressure changes.
The basis for determination of the low-low setpoint is the Loss of Normal Feedwater and Feedline BreaP events.
The setpoints.were determined by considering the level used in each of the analyses for each unit.
Unit 1 Unit 2.
O o
toss of Normal 0%
Feedline Break 2.
0%
49 For each unit, the setpoint was determined by considering the folicwing errors-fcr feedline -
break:
1
- Normal errors (normal channel accuracy, etc.)
- Post-Accident effects on transmitter i
(radiation and temperature)
- Reference leg effects (post-accident heatup)
L
\\
032-131 AMENDMENT 49 JUNE 5, 1984
_