ML20137M979
| ML20137M979 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 11/27/1985 |
| From: | Tucker H DUKE POWER CO. |
| To: | Harold Denton, Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8512040044 | |
| Download: ML20137M979 (42) | |
Text
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Duxe POWER GOMPANY P.O. HOX 33180 CHAHLOTTE, N.C. 28242 HAL H. TUCKER g
' " " " " * " (704) 373-453 mm... m o. .
November 27, 1985-1 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory-Commission Washington, D. C. 20555 l Attention: Mr. B. J. Youngblood, Project Director )
PWR Project Directorate No. 4 Re: Catawba Nuclear Station Docket Nos. 50-413 and 50-414
Dear Mr. Denton:
By letter dated September 10, 1985, Duke proposed an amendment to the Technical Specification for the reactor trip setpoint on lo-lo steam generator water level for Catawba Unit 1. This letter provides supplemental information with respect to the earlier submittal.
As a result of discussions with the NRC staff the following additional information is being provided.
- 1. The previous Justification and Analysis of No Significant Hazards Consideration discussed changes to the analysis of Loss of Normal Feedwater and Loss of AC Power analyses.
These analyses have been further revised to update the decay heat assumption rather than revise the auxiliary feedwater assumption.
- 2. FSAR pages which will be revised as a result of the pro-posal Technical Specification changes are attached.
- 3. The September 10, 1985 amendment request also requested that the proposed changes be incorporated into the proposed Catawba Units 1 and 2 combined Technical Specifications which were previously submitted on March 15, 1985. Because Catawba Unit 2 has Model D-5 steam generators vs. Model D-3 for Unit 1, the reduced low-low level trip setpoint is not needed for Catawba Unit 2.
Marked up pages for the Catawba Units 1 and 2 Technical Specifications are attached.
DOC P I 1
I '
'Mr. Harold R. D nton, Director
-November 27,~1985 Page Two
- This letter contains information which supplements that which was provided by my letter.of September 10, 1985. As such'no additional
~ license fees are necessary.
Very truly yours, c
.sb A Hal B.' Tucker
'ROS:slb Attachment cc: Dr. J. Nelson Grace, Regional Administrator U. S. Nuclear Regulatory Commission Region II" 101'Marietta Street, NW, Suite 2900 Atlanta, Georgia '30323 NRC' Resident Inspector Catawba Nuclear Station
'Dr. K. Jabbour Office of Nuclear _ Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Mr. Heyward.Shealy, Chief Bureau of Radiological Health S. C. Department of Health &
Environmental Control 2600 Bull Street Columbia, South Carolina 29201
JUSTIFICATION AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION The steam generator low-low water level trip function protects the Jreactor from a loss of heat sink in the event of a sustained steam /feedwater flow mismatch. Analyses were performed to justify lowering the programmed steam generator low-low level setpoint for Catawba Unit.1. This setpoint change along with the addition of a filter to the channel circuitry, will, help prevent unnecessary reactor trips-as a result of load rejections. This change will also prevent unnecessary actuation of the turbine-driven auxiliary feedwater pump due to spurious steam generator low-low level
' indications which result from " ringing" in the level transmitters.
-This benefit has been verified by observation of McGuire reactor trips-since-implementation of this setpoint modification.
To verify the acceptability-of.the proposed changes, Westinghouse has reanalyzed the Loss'of Normal Feedwater, Loss of AC Power, and Feedwater System Pipe Break transients, which rely on the steam generator low-low level reactor trip for protection. The responses of various system parameters to the above analyzed accidents are given in the attached FSAR pages.. -Results of these analyses l indicate that all applicable safety criteria-are met using the revised setpoint and the increased instrument delay time.
For the loss of Normal.Feedwater and Loss of AC Power analyses, it was necessary to revise the original FSAR assumption for residual decay heat in' order to compensate for the lower low-low level
-setpoint. For these two analyses, core decay heat generation is
-based on ANSI /ANS-5.1 1979, "American National Standard for Decay
' Heat Power in Light Water--Reactors," August 1979. The FSAR will be revised to reflect the new assumptions.
The Feedwater Malfunction accident is the only other accident analyzed in the FSAR which takes credit for a rector trip on steam generator low-low level. During this accident a malfunction is postulated which causes an increase in feedwater flow. When the steam generator level in the faulted loop reaches the high-high level setpoint, all feedwater isolation valves and feedwater pump
-discharge valves are automatically closed and the main feedwater pumps are tripped.'This prevents a continuous addition of feedwater and initiates a turbine trip. The analysis does not take credit for a reactor trip on turbine-trip. Consequently following core power stabilizes at a reduced level
- the turbine with consistent triplhe reactivity parameters assumed to maximize the initial increase in power. The reactor is tripped on steam generator low-low level if no action is taken by-the operator to terminate the reduced power operation. The revised steam generator low-low level setpoint will only delay the time of reactor trip at the reduced power level. Even assuming a delayed reactor trip, the DNBR limit is not exceeded at any time during the accident.
This evaluation has examined the impact of the proposed steam generator low-low level setpoint on the accident analyses performed in the FSAR. All accidents which take credit for a reactor trip on steam generator low-low level were analyzed. The results of these analyses indicate that all safety criteria are met using the revised setpoint.
r .
10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the proposed amendment would not:- -
(1) Involve a significant~ increase in the probability or consequences of an accident previously evaluated; or (2) -Create the possibility ofta new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
The proposed amendment does not involve an increase in the probability or' consequences of any previously evaluated accident.
The accident analyses have been reviewed and all acceptance criteria has been satisfied.
The proposed amendment does not create tha possibility of a new or different kind of accident than any previously evaluated since there will be no physical changes made to any plant system other than the reduction-of the trip setpoint and the addition of the filter to the channel circuitry.
The proposed amendment does not involve a significant reduction in a margin of safety. All applicable safety analyses have been reviewed ~and all acceptance criteria will be met with the revised setpoint.
The. Commission has provided guidance concerning the application of standards of no significant hazards determination by providing certain examples (48 FR 14870). This change is similar to example (vi).
For the reasons stated above, it is concluded that the proposed amendment does not involve significant hazards considerations.
I
- g. , TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
' TOTAL SENSOR ALLOWANCE ERROR E FUNCTIONAL UNIT (TA) (S) TRIP SETPOINT Z_ ALLOWA8LE VALUE Q
sa 13. Steam Generator Water -
s g a. Unit 1 17 14.2 1.5 >17% of span >15.3% of span from o Level Low-Low w
Trom 0% to 30% D% to 30% RTP*
RTP* increasing ncreasing linearly 40,0 linearly to o >99-f% of span M of span roml30%to100%RTP*
Yrom 30% to 100%
RTP*
( N'3
- b. Unit 2 17 14.2 1:5 y 117% of narrow 115.3% of narrow us range span range span
- 14. Undervoltage - Reactor 8.57 0 1.0 >77% of bus Coolant Pumps ->76% (5016 vol,ts) voltage (5082 volts) with a 0.7s response time.
- 15. Underfrequency - Reactor 4.0 0 1.0 >56.4 Hz with a Coolant Pumps Q55.9Hz D.2s response time
- 16. Turbine Trip _ -
- a. Low Control Valve EH , N.A. N.A. N.A. >550 psig Pressure >500 psig o
- b. Turbine Stop Valve N.A. N.A. N.A. >1% open Closure 11% open h g 17. Safety Injection Input
-4 from ESF N.A. N.A. N.A. N.A. N.A.
- h '
O "RTP = RATED THERMAL POWER ' : '
~
5 TABLE 3.3-2 (Continued) 9 g REACTOR TRIP SYSTEM INSTRWENTATION RESPONSE TINES I
FUNCTIONAL UNIT RESPONSE TINE E
g 12. Low Reactor Coolant Flow "
" a. Single Loop (Above P-8) < 1 second g b. Two Loops (Above P-7 and below P-8) <
nd
- 13. Steam Generator Water Level-Low-Low < onds
- 14. Undervoltage-Reactor Coolant Pumps .5 seconds
- 15. Underfrequency-Reactor Coolant Pumps < 0.6 second E
- 16. Turbine Trip e
- a. Low Control Valve EH Pressure N.A. 2 -
Y Turbine Stop Valve Closure ce
- b. N.A. $
- 17. Safety Injection Input from ESF N.A.
- 18. Reactor Trip System Interlocks o N.A. o .
- 19. Reactor Trip Breakers N.A.
I 20. Automatic Trip and Interlock Logic tL 4. g
\ .
n-
-r V
A ce
(#1
TABLE 3.3-4 (Continued) 9 '
g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS h -
SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWA8LE VALUE z
Q 8. Auxiliary Feedwater (Continued) .
, u, g c. Steam Generator Water g Level - Low-Low o 1) Unit 1 17 14.2 1. 5 > 17% of span > 15.3% of Trom 0% to ipan from 0% to
. 30% RTP 30% P increasing j increasing line Iv to l li > of span ,J f.3
@.0 %nearly ofto Trom 30% to 100%
span from 30% RTP ~
4 to 100% RTP .
u *
! > 2) Unit 2 17 14.2 1.5 1 17% of 1 15.3% of narrow l
w narrow range range instrument J, span span .
- d. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
- e. Loss-of-Offsite Power N.A. N.A. N.A. > 3500 V > 3200 V
- f. Trip of All Main Feedwater j Pumps N.A. N.A. N.A. N.A. N.A. m o
i g. Auxiliary Feedwater Suction Q Pressure-Low h
- 1) CAPS 5220, 5221, 5222 N.A. N.A. N.A. 1 10.5 psig > 9.5 psig 2
l
- 2) CAPS 5230, 5231, 5232 N.A. N.A. N.A. > 6.2 psig 1 5.2 psig -
! o 9. Containment Sump Recirculation .
o o
j a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. Q
! and Actuation Relays x n
l -
- b. Refueling Water Storage N.A. N.A. N.A. > 177.15 inches 1,162.4 inches
$ Tank Level-Low i Coincident With Safety j Injection See# Item 1. above for all Safety Injection Setpoints and Allowable Values.
l
Thio Um' "' W' '
I'N,g;,,,i n1.$vain .lsr B,C seconds . As rol clus+er cenirst aesr4Nes. The D
'innvien+^
Mi ceMy** wl c$ #^y '* f f u ,, sed. Ths a re 'e*nalysed hr vorisu vr< sons 4he 3 30m ,"" U N. ;, m,,;y. is. 2. c., is. 2. 7, m l: '; * "kir *i' i" 'N*ns w, -
Is. 2. r, end 15. 3. 2 asd Hre Ll obre.
o as a fumk a of power,,
g,y f g
en:!y:: ssuc st~ed kr lhese coemetents, Coolant Syst 1 as* loss of reactor coolant from cracks or ruptures in the R actor given in Ta le do not ug,15. 0. 3-2depend on reactivity feedback effects.
. The values re which showsgthe upper and lower bound Doppler ef pwer, used in the transient analysis. 0 functias power c The justification for use a of con-servatively event-by large event versus small reactivity coefficient values ar,e treated on an basis.
used to bound the effects of core life,In Jer e m some ala cases conservative combination 4ent it ;= conservateta "ea in a leu ince==ca trane-
-ef.Lictent_ alys5h +hes e condninnNen sa all de@ies defed and o a.meil ned rator ce u M*f nof refresent fess Eble runs &ie. Kihe*Nont, 15.0.5 na;,e v ers en RODclurnefer:sQ CLUSTER CONTROL ASSEMBLY INSERTION CHARACTERISTICS Hme +
he negative reactivity insertion following a reactor '
1 ip is a function of the as a functior,of rod position.:: cele.et'ca of tt.e rod cluster control assemblies an With respect to accident 1
parameter ir the time of insertion up to the dashpot ent .-er ses, the critical pproximately 85 percant o( e g od clu4ter travcl. The rod cluster contro
}
versus timeaassumed inaaccident analyses is shown in Figure 15.0.5-1. assembly The rod position cluster control assembly insertion time to dashpot entry"*4 taken as 3.05 second unle:: Otherw he noted in thc d h: = f =.
dependent en the type af c1"ete" centrol sesee i;e= a d ua+1y thered plant-inDrop time te
( - end eie specified in the plant Technical Specifications,, and 4. taite*al s-laring
.tes+ verifica 4+ ei+ke e sefoJy nalysis a ss e.*yaien is conser<<Jiv e. s Figure 15.0.5-2 shows the fraction of total negative reactivity insertion versus normalized lower regionrod position of the core. for. a core where the axial distribution is skewed to the region of the core can arise from an unbalanced xenonThis distribution.An curve ax j is used to compute the negative reactivity insertion versus time following a reactortriQ,hichisinputtoallpointkineticscoremodelsusedintransient analyses.
point kinetics core model.%4 bottom skewed power distribution itself is not input into the There is inherent conservatism in the use of Figure 15.0.5-2 in that it is based on a skewed flux distribution which would exist relatively infrequently. For cases other than those associated with unbalanced xenon distributions, signi-ficant negative reactivity would have been inserted due to the more favorable axial distribution existing prior to trip.
a/x The normalized rod cluster contr time is shown in Figure 15.0.5 .
assembly negative reactivity insertion versus from Figures 15.0.5-1 and 15. 5-2.The curve shown in this figure was obtained l lowing a trip of 4 percent A total negative reactivity insertion fol-specifically noted otherwise. is assumed in the transient analyses except where the calculated trip reactivity worth available as shown in Table .. For 4 3 2-3Thi Figures 15.0.5-1 and 15.0.5-2, the rod cluster control assembly drop time is nor-malized to 3.05 seconds, unless otherwise noted for a particular event.
The normalized rod cluster control assembly negative reactivity insertion versus time curve for an axial power distribution skewed to the bo . .
15.0-8 I
--k "'
Far h2 tots of non- emryertcy Ac power- Jo +h2 sh Hen audlia,;e s , SecNon 15.2 6 8
, , ne d +ke ten o4 normt ferdader , Seelion Is.2.7, core residget hed generSon la based a rt Rehrersee s . For +hs loss d ecohni acciden+,s nH:n 1s.4.5, resMaal I, ea f gene raHse.
CNS 15.0.10 RESIOUAL DECAY HEAT 15.0.10.1 Total Residual Heat
- n. ja"e! M et h a escritical ccre is calculated for the 10:s of ccclant =<-
-e4 dent per the requirements of Appendix K of 10CFR50.46 (Reference 6) as des-cribed in References 7 and 8. These requirements include assuming infinite irradiation time before the core goes subcritical to determine fission product decay energy. For all other accidents, the spee- models are used except that fission product decay energy is based on corafaverage exposu at the end of the equilibrium cycle.
- seche. ss /s. G. 5
- f*
- 15.0.10.2 Distribution of Decay Heat Following Loss of Coolant A cident During a loss of coolant accident, the core is rapidly shut down by void for-mation or rod cluster control assembly insertion, or both, and a large frac-tion of the heat generation to be considered comes from fission product de-cay gamma rays. This heat is not distributed in the same manner as steady state fission power. Local peaking effects which are important for the neu-tron dependent part of the heat generation do.not apply to the gamma ray con-tribution. The steady state factor of 97.4 percent which represents the frac-tion of heat generated within the clad and pellet drops to 95 percent for the i
hot rod in a loss of coolant accident.
C. For example, consider the transient resulting from the postulated double ended break of the largest Reactor Coolant System pipe; 1/2 second after the rupture about 30 percent of the heat generated in the fuel rods is from gamma ray absorp-tion. The gamma power shape is less peaked than the steady state fission power shape, reducing the energy deposited in the hot rod at the expense of adjacent colder rods. A conservative estimate of this effect is a reduction of 10 percent of the gamma ray contribution or 3 percent of the total. Since the water density is considerably reduced at this time, an average of 98 percent of the available heat is deposited in the fuel rods, the remaining 2 percent being absorbed by water, thimbles, sleeves and grids. The net effect is a factor of 0.95 rather than 0.974, to be applied to the heat production in the hot rod.
15.0.11 COMPUTER CODES UTILIZED Summaries of some of the principal computer codes used in transient analyses are given below. Other codes, in particular very specialized codes in which the modeling has been developed to simulate one given accident, such as those used in the analysis of the Reactor Coolant System pipe rupture (Section 15.6), are summarized in their respective accident analyses sections. The codes used in the analyses of each transient have been listed in Table 15.0.3-2.
15.0.11.1 FACTRAN FACTRAN calculates the transient temperature distribution in a cross section of a metal clad 002 fuel rod and the transient heat flux at the surface of the clad using as input the nuclear power and the time-dependent coolant parameters 15.0-11 i
\
chslemer, H , Bs em, l. . H . , Shop , D. R . , "Imfesved Thermal Design Procch ye wc A P -85 s,7 , " July .19 75
)
1 CNS
- i. <
REFERENCES FOR SECTION 15.0 1.
DiNunno, J. J. , et al. , " Calculation for Distance Factors for Power and Test Reactor Sites," TID-14844, March 1962.
2.
ORNL-4628 M. J. Bell, May "0RIGEN 1973.- The ORNL Isotope Generation and Depletion Code,"
3.
RSIC-OLC-38, "0RIGEN Yields and Cross Sections - Nuclear Transmutation and Decay Data From END F/B-IV," Radiation Shielding Information Center, Oak Ridge National Laboratory, September 1975.
I 4.
'l B:':,
" J. , "L'r:r.iu;;; ",icxide P. epartic; and .9 clee- AppWatinne", uava!
,eowiv C;x.issi;r., =, C;viei;n 1051. cf "::: tor 0 vclep. nt United 5Laica Atemic Energy -
h01 D-
- 5. -S d d,'^. F,, "A 1
Segge:t d tieti.ed fur Caiuulating use O!ffu;ien of Dadic-
-::tive "se N C;; Fi::ica Pr: duct: Fr : 'JO Fu 1 El:: r.ts," DCI-27 1957.
6.
" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10CFR50.46 and Appendix K of 10CFR50.
Federal Register, Volume 39, Number 3, January 4,1974.
- 7. Bordelon, F. M., et al., " SATAN-VI Program:
Comprehensive Space-Time Dependent Analysis of Loss of Coolant," WCAP-8302 (Proprietary), and WCAP-8306 (Non-Proprietary), June 1974.
- 8. Bordelon, F. M. , et al. , "LOCAT-IV Program: Loss of Coolant Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary),
June 1974.
- 9. Hargrove, H. G. , "FACTRAN - A Fortran-IV Code for Thermal Transients in a U02 Fuel Rod," WCAP-7908, June 1972.
- 10. Burnett, T. W. T.,
et al., "LOFTRAN Code Description," WCAP-7907, June 1972.
11.
Risher, D. H. , Jr. and Barry, R. F. , " TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code," WCAP-7979-P-A (Proprietary), and WCAP-8028-A (Non-Proprietary), January 1975.
- 12. "'l:vinghn"se Nucia;;- En ;gy Syste-e n<u< sten qus;itj A33urence Pleis,"
WCA" 0370 A.~
ANs1l A NG- s. I- l171, " Arner,'enn NaNenal S+anb ral Sr bery Nee 4 Power- for figh+ We,hv Reaclers", Auys + /1 '71.
l 15.0-14 '
.o. .n._ ,u
No l e. J h d Jfia one me reylaces hon p 4j e s. of Me o/d Table 15.0.6-1 ff/6 / 5 . O. (, - / Trio Points and Time Delays To Trio Assumed In Accident Analyses Trip
- Limiting Trip Point Assumed Time.Delaps Function in Analysis (Seconds)
Power range high neutron flux, high setting 118% 0.5 Power range high neutron flux, low setting 35%
0:5 High neutron flux, P-8 85% 0.5 Overtemperature AT Variable see 6.0*
Figure 15.0.3-1 Overpower AT Variable see 6.0*
Figure 15.0.3-1 High pressurizer pressure 2410 psig " 2.0 litt low pressurizer pressure 4636- p s ig * *" 2.0 Low reactor coolant flow '
(from loop flow detectors) 87% loop flow , 1.0 Undervoltage tr'ip 68% nominal 1.5 Turbine trip Not applicable 2.0 Low-low steam generator level
- r. 4.0 High steam generator level nc%
67-4% of narrow range 2.0 feedwater pump trip, level span *"'"
feedwater isolation, and turbine trip
- Total time delay (including RTD bypass loop fluid transport delay effect, bypass loop piping thermal capacity, RTO time response, and trip circuit, channel electronics delay) from the time the temperature difference in the coolant loops exceeds the trip setpoint until the rods are free to' fall.
- The num er', cal s e+(oin t assymed. for #is + rip kneHon varies deper,d;ag on %e acciolen+ being y ausa l ,.ed.
The values used are given in. +lne descrifrip lia,1 eelee.to., snn g earraler 1,,,t, ats,;+. E*t s of 4he various acciden aos ca m wl,acit resul+ in *- .
rf,*c, ,r t is ,,, ore t;,,, % ,3 a ,.x ;s nas asex 4, the
- We olis ere a ncir.s between the values and 4 hose gl vert +ke the
.%);en s Ca+ a w ha Alaclear e;.+ Melhedoley ahearnent are nddressed in s 7mne .ts,IttY lettw %
H.B. neker D&e Power) +e R. E. benk (NAC)
% . ii
CNS even if a loss of all non-emergency AC power occurs simultaneously with loss of
()
- normal feedwater. The turbine exhausts the secondary steam to the atmosphere.
The pumps take suction from the auxiliary feedwater storage tank for delivery to the steam generators.
Upon the loss of power to the reactor coolant pumps, coolant flow necessary for core cooling and the removal of residual heat is maintained by natural circu-
, lation in.the reactor coolant loops.
- Aloss$fno'n-emergencyACpowertothestationauxiliariesisclassifiedas En ANS Condition II event, fault of moderate frequency. See Section 15.0.1 for a discussion of Condition II events.
A loss of AC power event, as described above, is a more limiting event than the turbine-trip-initiated decrease in secondary heat removal without loss of AC power, which was analyzed in Section 15.2.3. However, a loss of AC power to the plant auxiliaries as postulated above could result in a loss of normal feedwater'if the condensate pumps lose their power supply.
Following the reactor coolant pump coastdown caused by the loss of AC power, the naturaVcirculation capability of the RCS will remove residual and decay heat from the core, aided by auxiliary feedwater in the secondary system. An analysis is presented here to show that the natural circulation flow in the RCS following a loss of a-c power event is sufficient to remove residual heat from the core., )
The plant systems and equipment available to mitigate the consequences of a loss of AC power'ovent are discussed in Section 15.0.8 and listed in Table 15.0.8-1.
, 15.2.6.2 Analysis of Effects and Consequences Method of Analysis A detailed analysis using the LOFTRAN Code (Reference 3) is performed to obtain the natural circulation flow following a station blackout.' The simulation describes the plant thermal kinetics, Reactor Coolant System (RCS) including the natural circulation, pressurizer, steam generators and feedwater system.
The digital 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 design rating. 1
- 2. ^ ::n;;rvative core red @e! 5ert generetf e- 5 ::d upon len; t: pm auen j at the "itfe! ~~e-leve! preceding the trfp. ;
j I
' 3. A heat transfer coefficient in the steam generator associated with RCS natural circulation.
Tcore reManal hea+ gener*Nen is based o'1 4he (1 74 version of ANS-5l (Reference Q.
L This nAod is
~
a co nserveliv e refresenl4lon o-( Jhe ciecay enemy release rates.
15.2-10
CNS #'
g e* m\ U gsumed ho beM, a+ f* N *" eyt.
C 4. Reactor trip occurs on steam generator low M ievelj No credit is taken for immediate release of the control rod drive mechanisms caused by a loss of offsite power.
5.
The worst single failure in the auxiliary feedwater system occurs. 49I ym o huxiliary feedwater is delivered to two steam generators.
{ . Secondary system steam relief is achieved through the steam generator safety valves.
The assumptions used! i n the analysis are essentially identical to the l'oss of normal feedwater flow incident:(Section 15.2.7) except that power is assumed to be lost to the reactor coolant pumps at the time of reactor trip. '
Plant characteristics and initial conditions are further discussed in Section 15.0.3.
Results .dr,g 15. 2. 4 -4 The transient response of the RCS following a loss of ac power is shown in l Figures 15. 2. 6-1c15. . C-- 2, and 15.2. 0 3. The calculated sequence of events for this event is listed in Table 15.2.3-1.
lThe first few seconds after the loss of power to the reactor coolant pumps C 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 tempera-tures 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.
The LOFTRAN code results show that the natural circulation flow available is sufficient to provide adequate core decay heat removal following reactor trip and RCP coastdown.
15.2.6.3 Environmental Consequences The postulated accidents involving release of steam from the secondary system will not result in a release of radioactivity unless there is leakage from the Reactor Coolant System to the secondary system in the steam generator. A con-servative analysis of the potential offsite doses resulting from this accident is presented assuming primary to secondary leakage. This analysis incorporates assumptions of 1 percent defective fuel and existence of a 1 gpm steam generator leak rate' prior to the postulated accident for a time sufficient to establish equilibrium specific activity levels in the secondary system. Three postulated cases are analyzed:
Case 1 (No iodine spike)
Case 2 (With prei existing iodine spike)
Case 3 (With coincident iodine spike) 15.2-11 -Rcv. / /*
CNS 15.2.7.2 Analysis of Effects and Consequences Method of Analysis A detailed analysis using the LOFTRAN Code (Reference 3) is performed in order to obtain the plant transient following a loss of normal feedwater. The simu-lation describes the plant thermal kinetics, RCS including the natural circu-lation, pressurizer, steam generators and feedwater system.
computes pertinent variables including the steam genentor level, pressurizerThe digital p water level, and reactor coolant average temperature.
Assumptions made in the analysis are:
1.
The plantdesign Features is initially operating at 102 percent of the Engineered Safety rating.
2.
- A conservative core residual heat generation based upon long term operation at the initial power level y ced g the trip. f 4g ,.y 3.
.g Reactor trip occurs on steam generator low-lo? level, assum/ Jo be a+ ts: set, 4.
f 'The worst single failure in the auxiliary feedwater system occurs. f M biliary feedwater is delivered to two steam generators.
$ )(. Secondary system steam relief is achieved through the steam generator i
safety valves.
The loss of normal feedwater analysis is performed to demonstrate the adequacy Feedwater System) in removing long term decay heat and prev heatup of the RCS with possible resultant RCS overpressurization or loss of RCS water.
As such, the assumptions used in this analysis are designed to minimize the energy removal capability of the system and to maximize the possibility of water relief from noted in thethe coolant system assumptions listedbyabove.
maximizing the coolant system expansion, as For the loss of normal feedwater transient, the reactor coolant volumetric flow remains level trip. at its normal value and the reactor trips via the low-low steam generator The reactor coolant pumps may be manually tripped at some later time to reduce heat addition to the RCS.
An additional assumption made for the loss of normal feedwater 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 psia) throughout the transient.
l Plant characteristics 15.O.3. and initial conditions are further. discussed in Section 15.2-14 b
i
-Rev. X /b f
C CNS Plant systems and equipment which are necessary to mitigate the effects of a loss of normal feedwater accident are di: cussed in Section 15.0.8 and listed in Table 15.0.8-1. Normal reactor control systems are not required to function.
The Reactor Protection System is required to function following a loss of nor-mal feedwater as analyzed here. The Auxiliary Feedwater System is required to deliver a minimum auxiliary feedwater flowrate. No single active failure will prevent operation of any system required to function. A discussion of ATWT con-siderations is presented in Reference 2.
Results
,y .
+hre U. 2. '7-Y3 Figures 15.2.7-415.2.f'-2, 2nd 15.2.7-3 show the significant plant para 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 fract-tion and because steam flow through the safety valves continues to dissipate the stored and generated heat. One minute following the initiation'of the low-
~
~
low level trip, at least one auxiliary feedwater pump is automatically star-ted, reducing the rate of water level decrease.
l l The capacity of th 4 xiliar dwate fy tem is such that the water level in the steam generator being f W does not ecede below the lowest level at which sufficient heat transfer area is available to dissipate co esi 1 heat I
without water relief from the RCS safety valves. .5.2.'-1 sh ws that at no time is there water relief from the pressurizer. Figure (/5.2 .7-2j/
The calculated sequence of events for this accident is listed in Table 15.2.3-1.
As shown in Figures 15.2.7-1 and 15.2.7-2, the plant approaches a stabilized condition following reactor trip and auxiliary feedwater initiation. Plant procedures may be followed to further cool down the plant. ,
15.2.7.3 Environmental Consequences If steam dump to the condenser is assumed to be lost, heat removal from the sec-ondary system would occur through the steam' generator power relief valves or safety valves.
Since no fuel damage is postulated to occur, radiological. con-sequences resulting from this transient would be less severe than the steam-line break accident analyzed in Section 15.1.5.3.
15.2.7.4 Conclusions Results of the analysis show that a loss of normal feedwater does not adversely affect the core, the RCS, or the steam system since the auxiliary feedwater capacity is such that reactor coolant water is not relieved from the pressurizer i relief or safety valves. The rcdiclogical cancan"an' ac af+* ^"^^+ " " "
=100: Ocycrc than the Otccali.7 brc k accidea+ a naly7ari in < a+ 4 a a F 1.5.2.
15.2-15 a av. f /:/-
/
l CNS 4.
Initial pressurizer level is at the nominal programmed value plus 5 per-cent (error); initial steam generator water level is at the nomi e l plus 5 percent in the faulted steam generator and at the nomina ve4*e inus 5 percent in the intact steam generators. Va ld e 5.
No credit is taken for the high pressurizer pressure reactor trip.
- 6. Main feedwater flow to all steam generators is assumed to be lost at the time the break occurs (all main feedwater spills out through the break).
- 7. The worst possible break area is assumed.
This maximizes the blowdown dis-charge rate following the time of trip, which maximizes the resultant heatup of the reactor coolant.
, 8. owHI A conservative feedline break discharge quality is assumed rior to the l 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 genera-t'or. This minimizes the heat removal capability of the affected steam gen-
'erator.
- ,;;; Q a+ s7 (enee of urrow me sjoen belo
_l. 9. Reactor trip is assumed to be initialed the low-low level trip set-point =inus 10 perccat of t.m ur .~. ca;;; cpam i the faulted steam gen-erator.1: reached.
adped
- 10. The Auxiliary Feedwater System is actuated by the low-low steam generator water level signal. The Auxiliary Feedwater System is assumed to supply a I
total of 492 gallons per minute (gpm) to the two unaffected steam gen-erators, including allowance for possible spillage through the main feed-water line break. A 60 second delay was assumed following the low-low level signal to allow time for startup of the emergency diesel generators and the auxiliary feedwater pumps. An additional 115 seconds was assumed before the feedwater lines were purged and the relatively cold (134 F) aux-iliary feedwater entered the unaffected steam generators.
- 12. No credit is taken for charging or letdown.
- 13. Steam generator heat transfer area is assumed to decrease as the shell side liquid inventory decreases.
- 14. Conservative core residual heat generation is assumed based upon long term operation at the initial power level preceding the trip.
15.
No credit is taken for the following potential protection logic signals to mitigate the consequences of the accident:
I 15.2-18 nav. ! /N
CNS (l a. High pressurizer pressure.
- b. Overtemperature AT.
- c. High pressurizer level.
- d. High Containment pressure.
Receipt of a low-low steam generator water level signal in at least one steam generator starts the motor driven auxiliary feedwater pumps, which then deliver auxiliary feedwater flow to the steam generators. The turbine driven auxiliary feedwater pump is initiated if the low-low steam generator water signal is reached in at least two steam generators. Similarly, receipt of a low steam line pressure signal in at least one steam line initiates a steam line isola-tion signal which closes the main steam line isolation valves in all steam lines. This signal also gives a safety injection signal which initiates flow of borated water into the RCS. The amount of safety injection flow is a func-tion of RCS pressure.
Emergency operating procedures following a secondary system line rupture call for the following actions to be taken by the reactor operator:
- 1. Isolate feedwater flow spilling out the break of ruptured steam generator and align system so level in intact steam generators recovers.
- %e
- 2. op Af gh he&d safety injection -charging pumps if:
9 s+able or hscreaso,ng.
- a. Wide range reactor coolant pressure is greater ther 200n neig, C -and h st2Ll Or Scre2 9 3-
- b. Pressurizer water level is greatcr than 50 ;'$r" cent er span.
- c. RCS is adequately subcooled. ,
- sa friele.& **
- d. Steamgeneratornarrow[angelevelindicatinexistsinatleast one steam generator or%uxiliary feed ter eing injected into -et /fe t;;st On: n0r-faulted steam generat rs /, ge,.,/de a,, =4 7 5. hea/- sink.
I Subsequent to recovery of level in the intact steam generators, the-high h:21 aafety injectie.. pumea will be turned of' and plant operating procedures will be followed in cooling the plant to hot shutdown conditions.
Plant characteristics and initial conditions are further discussed in Section 15.0.3.
No reactor control systems are assumed to function. The Reactor Protection System is required to function following a feedwater line rupture as analyzed here. No single active failure will prevent operation of this system. A sis-l e ,usio n d Arwr canaider.Ha ns is pres e,,pex in ge ference. 2.
The engineered safety systems assumed to function are the Auxiliary Feedwater System and the Safety Injection System. For the Auxiliary Feedwater System, l the worst case configuration has been used, i.e. , two intact steam generators receive auxiliary feedwater following the break.70ne motor driven auxiliary ;
r ;
ne +uri,ine driven miliary fee.(ader pusy is usumeel J, fail. n 15.2-19 -Rcv. T /F
_CNS
, ;}) i+s enFire flow o & +he break.
l feedwater pump has been assumed to ,T i
juiththeturbir.: 9 4 'f e a p" p delivers second motor driven pump t gether ators.ellowir.; for :p*'hg cut of th: br::k. to the two intact steam gener-gpm Only one train of safety injec-tion has been assumed to be available.
For the case without offsite power there will be a flow coastdown until flow in the loops reaches the natural circulation value.
ability of the RCS has been shown in Section 15.2.6, for the loss of AC non-The emergency power transient, to be sufficient to remove core decay heat following t
f i
f i
t 15.2-19a "ev. 3 ! _ce. , nuar .
I I
e
CNS REFERENCES FOR SECTION 15.2
'l 1.
Mangan,M.A.,f.erpressureProtectionforWestinghousePressurized Water Reactor WCAP-776_9, October 1971.
2.
" Westinghouse Anticipated Transients Without Trip Analysis",
'WCAP-8330, August 1974.
3.
Burnett, T. W. T. , et al. , "LOFTRAN Code Description," '!CA" 7^07,
-Ja..: 1972.
- 4. Hargrove, H. G "
FACTRAN-A Fortran-IV Code for Thermal a UO2 Fuel Ro f' WCAP-7908, ansients in June 1972.
l- 5. Lang, G. E., Cunning J. B., " Report o onsequences of a Postulated J Main Feedline Ruptur WCAP _ , anuary 1978.
Apr*l, I18'/.* wcAP-7107-P-AlPropen'e4eq)and WCAP-7 , y Q
G.
' Mar lAns -6.l-1971, Amers'can NaWwal 9andard kr Deeg He-9 fower o'n
., u y0A4 r/aler Reaclors# Aug u s+ 1977.
3 1
i 15.2-22 w , y --,,ry- - vv m , __n--w --- - . - - ------- , . . - ~g- y- , --. _-_%- - - , - - - - - - , . - - - - , . - . ,
~
5 4 -
g-Table 15.2.3-1 (Page y 36-@ (,
Time Seouence Of Events For Incidents Which Cause A Decrease In Heat Removal By The Secondary System Accident Event Time (sec)
Loss of Non-Emergency Main feedwater flow stops 10 N
Low-low steam generator Of)( 54
, level reactor trip '
il Rods begin to drop SK.A( 58 ll Reactor cool pumps )( K 'o begin to coa wn .
Il
, Peak water level in IMC g '2
!f pressurtzer occurs A. .it.,7 f...I.,.4 e r r= f. st=rt M ##
f( T-* -Fette steam generators 125.,M('f8 begin to receive auxiliary feedwater from tiid motor driven ux'liery feeJ-oicr I
ll Core decay heat decreases ~ N478 1500 to auxiliary feedwater heat r,emoval capacity i
j[ Loss of Normal Main feedwater flow stops 10 Feedwater Flow Low-low steam generator level reactor trip
)()( 5'l Rods begin to drop ,K M 58 il p I Peak water level in n+C.
g u pressurizer occurs '#
- "1 a uia,y cre t +<r r.-p set T*o -Feter steam generators t M'.-fef /90 begin to receive auxiliary' feedwater from -t/AI motor-( driven cuxiliary fccc.;;ter -
Core decay heat plus pump ~,4096 ,%Ffift 3soo heat decreases to auxiliary feedwater heat removal capacity
_ & an A
. 1 l
(
l 1
X Q
Table 15.2.3-1 (Page X ef-5) j Time Seouence Of Events For Incidents Which Cause A Decrease In Heat Removal By The Secondary System Accident Event Time (sec)
Feedwater System Pipe Break l
- 1. With-offsite power Main feedline rupture occurs
, available )6' 10 Low-low steam generator level JK W 2P reactor trip setpoint reached in ruptured steam generator
~
Rods begin to drop Ff X 31 Mwary seeM e p..p aar+ y 87 Two sienn generalor.s "''i;ry fe&daatcr i ^ N begin fa receive aupLry deli','cred tc int 2ct 5% f Terdwater fram one nedar S tc c.T gencretes a driven ymp Low steam line pressure setpoint reached in rup- $% 244 tured steam generator All main steam line M y 253 isolation valves close l Steam generator safety valve setpoint reached M% 40/
in intact steam generators .
Pressurizer water relief 188E.f71 /V'/
begins .
, Core decay heat plus pump M) pac +Yv04 heat decreases to auxiliary feedwater heat removal
. capacity .
e t
9
- lS ' 'Jp de t c -
M
't -+-
Table 15.2.3-1 (Page 4 of-Ef)
Time Secuence Of Events For Incidents Which Cause A Decrease In Heat Removal By The Secondary System Accident Event Time (sec)
Feedwater System Pipe Break
, 2. Without offsite power Main feedline rupture occurs is 10 Low-low steam generator X >f ze level reactor trip set-point reached in ruptured steam generator p 32.
{Rodsbegintodrop)her jMKK W lost to the reactor cool-ant pumps Two sfeam genera 4e r.s - Antliaq 4eedwaer rums s+ art L,y4143.y f::d,;;t;r b -
Y *1 begin to recrove ouillery -dz p;;;a w 7,tggt
- K sK uy Jeedvaler bm one ess/pr _Ste2 gcr,;ratsc3 ofriren furep '3
' Low steam line pressure 2df.P(k 2#
setpoint reached in rup-tured steam generator i All main steam line 4rg2V 25.5 isolation valves close Steam generator safety valve setpoint reached MM 53#
in intact steam generators Core decay heat decreases to auxiliary feedwater K 2006 Ta4 */Sto
, heat removal capacity .
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Reactor Coolant Temperature Transients for the Faulted and the intact Loops for Main Feedline Rupture with Offsite Power Available CATAWBA NUCLEAR STATION .
Figure 15.2.8 3
14429 10
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Steam Generator Pressure and Core Heat Flux Transients for Main Feedline Rupture with Offsite Power Available CATAWBA NUCLEAR STATION Figure 15.2.8 4
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Nuclear Power Transient, Total Core Reactivty Transient, and Feedline Breakflow Transient for Main Feed!!ne Rupture without Offsite Powar Available CATAWBA NUCLEAR STATION '
Figure 15.2.8 5
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Reactor Coolant Temperature Transients for the Faulted and intact Loops for Main Feedline Rupture without Offsite Power Available CATAWBA NUCLEAR STATION Figure 15.2.8-7
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s 250 -
FAtJLTED STEAM
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N 8
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8 g i l l lllll l l I lllll 'i ;:'!!" ' ' I l!Ill I 5 to 5 10 2 5 10 3 5 104 TIME (SEC)
Steam Generator Pressure and Core Heat Flux Transients for Main Feedline Rupture without offsite Power Available CATAWBA NUCLEAR STATION Figure 15.2.8-8
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