ML19241A938
| ML19241A938 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/02/1979 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Ippolito T Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7907110452 | |
| Download: ML19241A938 (28) | |
Text
.
TENNESSEE VALLEY AUTHCRITY CH A TT A NCC G A. TEN'.E3SE E 37'~1 400 Chestnut Street Tower II July 2, 1979 Director of Nuclear Reactor Regulation Attention:
Mr. Thomas A. Ippolito, Chief Branch No. 3 Division of Operating Reactors U.S. Nuclear Regulatory Cocmission Washington, DC 22555
Dear Mr. Ippolito:
In.ne hit av of the
)
Docket Nos. 50-259 Tenres-e Valle; 1ithority
)
50-260 "0-296 i'is..i f t. response to A. Schwencer's letter dated January 13, 1978, x;ning suppression pool temperature transients at Browns Ferry Nuclear
~m c ia. ?.
As you are avare, we are deeply involved in the Mark I Long-Term Program (LTP) for which you indicated that the requested information will serve as part of the basis for your review.
We have examined the five suppression pcol temperature transient analyses requested, part A, 1(a) through (e),
and note that these are primarily concerned with the performance of SRV discharge through the existing Ramshead devices. Presently, it is our intent as part of the long-term program solution to replace these devices with "T"-Quenchers.
Accordingly, we have selected for analysis only those transiente which are the moat limiting.
By this approach we are providing you with the necessary information to deconstrate the satisfactory and conservative design of the Browns Ferry Nuclear Plant while avoiding severe impact on our LTP analysis and modification effort and schedule. provides our analysis primarily of your cases part A, 1(a) and 1(b). Although we did not specifically address cases 1(c) and 1(d), we have analyzed the case where two additional valves are opened above 120 F pool temperature with no heat exchangers in operation. This is bounding for cases 1(c) and 1(d). contains the requested information (part A,2) concerning the Browns Ferry Nuclear Plant suppression pool temperature monitoring system. Results described in Enclosure 1 are also discussed here. discusses the conservatism of the analysis presented in Enclosure 1 in light of the theoretical sequence of events and initial conditions.
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Mr. Thomas A.
Ippolito July 2, 1979 It is our understanding that the information requested in part B of your letter has been supplied on a generic basis in a September 1977 letter f rom E. D. Fuller, General Electric, to Olan D. Parr, Chief, LWR Branch No. 3.
This information should be made a part of the Browns Ferry Nuclear Plant dockets.
If your staff has any questions regarding the enclosed material, please get in touch with us.
Very truly yours, TENNESSEE VALLEY AUTHORITY
[
. M. Mills, Manager Nuclear Regulation and Safety Enclosures a
47%
E; CLOSURE 1 O POW:IS FEBRY !;UCLEAR PLA!!T SUPPPESSIO!1 FOGL At1ALYSIS Doscription of "emperature Transient Peculting Frcm A Stuck Open ' a fety Relief Valve J
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Tablo of Cant ent s I.
Introduction II. Analysis Description III. Assumptions I '1.
Results and Conclusions V.
9eforences 9
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I.
Introduction The Gen"ral Electric (GE) Mark I containment concept used at t l' o Erowns Ferry nuclear plant employs a torun suppression pool design as an in te rm ed ia te h"at sink during normal and accident conditions.
The subcooled wator in the pool serves a dual role in the Mark I containment system.
P rima rily it functions to limit containment pressure in the unlikely event of a lous-of-coolant accid en t by thermodynamically abscrbing the energy released in the form of steam.
Similarily and socondarily, the pool is designed to accommodate the main stean relief line discharge during normal plant operation.
It is this latter function that will be addressed by this analysis.
Concerns have recently developed that unstable steam conden sation at the main steam relief li ie and suppression pool interface may occur during relief valve discharge at elevated pool temperatures.
The condensation instability results in pool prescure oscillations and relief line vibraticns which are trar.smitted to tho torus shell resulting in unacceptably large st ru ct ural loadings.
Thin condensation phenomonon is not completely un?.erstood at the present time, however, conditions favorable to the instability occur when high steam flowrates are concurrent with high pool temporatures.
Therefore, GE has recommended an upper limit en
- he torun pool tempe rature for high SPV mass fluxes.
The tomperature and mass flux criteria suggested for the ramshead discharge device typical of the Browns Ferry design are 1600 F
( loca l), 1500 F (bulk) 1 for mass fluxes greater than 40 lt /sec ft2
~ho Nuclear Fegulatory Commission has roquested that all utilities with Mark I containments pe r f o rn a plant unique analysis to demonstrate that the GE criteria is not exceeded during a transient resulting from a stuck cren safety relief valve.
This report summarizes the conse rvative analysis used to examine *his pool temperature problem for the Browns Ferry nuclear plant.
1 "'. o r ul? root terroraturn ta the manu Tverage t o ru a pool m
vre,*ure, whernas *he local t q erature in confine! ta a few pipe diametera from the dinctIrao levice.
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Analysis Description Th e stuck open relief valve transient was examined using a combination of hand calculations and two computer models.
Initially, a hand calculation was used with the simplifying accumptions listed in the next section, to determine the time required to heat the suppression pool water to the technical specification limit for reacto r scram.
The calculations were then performed by computer analysis due to the complex system interactions following reactor scram.
7tn existing PETRAN2 computer molol of the Browns Ferry plant was modified to include the torus pool, PHP piping and the PHP heat exchangers.
Several control system models, including the feedwater csntrol system are included which permits feedwater mmlulation.
The PETRAN model was used until 50 seconds tollowing the scram.
At this point the Main steam Isolation Valves MSIV's had been closed and the system was isolated.
Eeyond this point a simplified program was written to iteratively balance the system energy inputs and losses over small time steps.
This technique was used to limit computer costs associated with achieving the required low ramshead mass fluxes using the more elabora te RETRAN code.
This code contains all the major energy inputs including feedwater, decay heat, sensible heat from core steel, coolant inventory, and recirculation system piping as in FETRAM.
An energy equilibrium is assumed to exist over each tire step such that the energy lost by blowd own is equal to the summation of all energy inputs.
Since the sensible heat terms depend on the vessel pressure which in turn depends on vessel blowdewn, an iterative technique is employed.
The feedwater inlet flow is balanced to the blowdown flow resulting in a net change in core water level of zero.
The docay heat curves w:re taken from NFC branch technical position ASE9-2 Fev 1 assuming fission product decay uncertainty factors of 1.2 before 1000 seconds and 1.1 thereafter.
Heavy element decay heat and an infinite operating tire were used to be conservative.
Figure 1 illustrates the core power versus tire used in the analysis.
It should be noted that on this plot and others, the initial 400 seconds of time required to heat the pool to 1100 F at a constant power are not shown.
2PFT9AN is the DELAP4 based computer code :leveloped by EPPI for operational transient simulation (1).
III. Assumptions covoral simplifying and conservative assumptions have toen made in this a na ly n i s.
These assumptions aro liuted and lescribed in this section.
T1.e pool tempe rature transient is drivon by the blowdown energy of the reactor.
In order that this energy is maximized, the reactor is considered to be initially at a steady state of 103 percent power.
After the reactor scram, the conservative for of the decav trat equation proviousiv described is used, resulting in an uppor bound on reactor pos. throughout the transient.
upprossion pool parameters are also rounding.
The pool is o
initially at its maxinun technical specification temperature of 950 F and its nininum water volume.
This maxiT.izes the initial energy content of the subccoled pool water within the constraint ot plant operating limits.
"'w o separate trains of residual heat removal system heat orchangars are ava ilacio to rerove energy from the torus pcol.
Each train consists of two rumps with a design flow rato of 10,000 grm and two heat oxchangers with heat transfer coefficients of approximately 270 Btu /sec o F.
The system is initiatod as prescribed by the plant technical spec i f ica t ions when the pool tongerature is 950 F and rojects heat to service water at 990 F.
Significant heat
- rans for does not occur until the pool temperature rises considerably above its initial value.
mho relief valve fails open at time zero and remains full.
- pen durinq the ntudy.
The flowrate selected for the valve is
- 1. 2 2 S t inos 'lo ASME ra ted flow for the valvo.
This f low ra te is obtained ii the PETRAL portion of the analys is bv so lect ing MCC"Y c ri - ica l flow and applyino the arpropriate crit ical flew con tra ct io n coe f ficient.
In the simplified program the flow is directly calculated.
During he 'io V transient, the pressure in the vessel would initially decroaso duo to the la r ger energy removal rate not ived iat oly accommodated b y the roactor system.
The turbine ontrol valvos automatically adjust during this time in an 9tompt to maintain the reactor stean dome pressure.
The unsol pressure cartially recovers duo to this action.
N ever, to simplify the a na lys is, this transient is ignore !
anl it is assumed that the initial pressure in the vessel is aintained through this period.
%e reactor vstem lumps steam to tho pool at tull powor unt.1 *he nool reacheu li0o F at which time the reactor i. s crammod by operator action as requiro1 by tho t ec hn ic a l rocifications.
"'his action occurs approximately 400 mennds int o t!.o t ra ns ie nt.
Conservatively, the r.a i n st oan i"olation valvos are usod in this nimulation to isolato *he
" ore.
The McIV's are assuned to remain cnen until a core low wator lovol signal is rocoivoi.
Minim'ization of this
- i-n and conroquently the energy lan* to the 'urbine is
'~*ainod Ev clo^ ins the fr e dwate r input to tho vessel at the
- ~ of scrar and rein st r it.-
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aro clonod.
Thic operation m win i z n the onorgy released to the pool since the core is "botttel u;" at aW
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higher decay heat power level.
- o energy is lost to the condenser via the turbine hypass system since thece valves are lef t arbitrarily closed throughout the simulation.
The pool temperature is instantly averaged through the torus at each time step in the simulation and therfore represents the bulk tempe rature.
The analysis is terminated once the ramshead mass flux reaches 40 lb /see f t2
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SUMMARY
OF ANALYSI'; PAPAMETERS (Base Case)
Reactor Power 103% full power (339 2 Mwt)
Reacrar Pressure 1035 psia SRV Flowrate 122.5% ASME flow (270 lb/sec initial)
Torus Pool Temperature 950 F (initial)
Scram Temperature 1100 F Torus Indication Decay Heat Infinite Irradiation & Heavy Elements No. of RHR Heat Exchangers 2
R HR Pump Flow 10,000 gal / min Heat Trans fer Coe f ficient 270 Etu/sec 0F Ramshead Device 10 Inch Schedule 80 Pipe
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IV.
Pesults and Conclusions Several studies were performed to de te rmine the options for achieving a low ERV mass flux without axceeding the pool temperature limit.
The options examined can be classified into two groups:
1.
Cpening of additional SRV's to promote a more rapid veccel depressurization.
2.
Use of addit ional heat exchangers to cool the suppression pool.
A tase case consisting of one valve t'
- stuck open valve) and one train of heat exchangers (2 exchangers) was selected since this case requires no operator action beyond scramming the reactor and assuring the RHR system is operating.
- he reactor pressure requirenont for the realization of 40 lb/sec ft2 ranshead mass flux was determined to be approximately 149 psia.
Below this pressure the mass flux is acceptably low for elevated pool temperatures.
Sincc the "norgy inputs from sensible heat depend primarily on the initial an, final vessel pressures (temnerature s), the time to achieve the reduced pressure only a r m ects the energy input from decay heat.
- herefore, a '. aster pressure decay would result in less energy transf or to the pool.
The base case pressure decay is shown in ficure 2.
The ra pid depressurization effect was examir.d by opening additional
= lief valves as shown in fi gu re s 3 end 4 wnere reactor scram occurs at tin +
0 and a pool temperature of 1100 F.
=
"he base case is represented by the 1 valve curve and as shown enters the condensation instability region.
- Mcwever, the use of additional valves opened 10 ninutes after scram (16 minutes after the valve sticks open) results in acceptable pool temperature behavior.
It should he noted that the temperature curves in Fig
- re 4 are terminated at the coint where the ramshead
- m. ass lux falls below 49 lt/sec ft2, thereby giving an ind ica t io n o' the time roquired to achieve that state.
Each additional valve opened has less effect than the previous resulting in no ra rticular advantage to opening more than 2 additional valvos.
Examination of the valve opening time requirements was perf rmed by comparing the base case to a situation where one additional valve was opened at various times following the reactor scran.
Fecults of this analysis, nhown in figures 5 and 6 indicate the opening time can be delaved until 1100 seconds a f ter scram.
Avoidance o f the condensation instability region can therofore he achieved
- hrough the uso of at least one additonal SR7 prior to 18 minutes after scram.
The operation of add it ion,1 hea t exchangers can also prevent
'ho entry into condensation instability.
rigures 7 and 8 indicate tho ope ra t ion of 3 or more heat exchangers fr*nuiring both t ra i ns) provides sufficient cooling capacity to torniaate the pool temporature rise prior to reachina 1500 F e ve n if no additional valvea are uned.
In each case
' worst case seconda ry aide temporature of 95o F was used to QT
ccnser/atively minimize the heat trr.sfer per heat exianger.
Censcrsely, a study was perforned, dccinented by Figures 9 and 10, *.o examine the cptions available if no Excol ccoling is available. The use of two additicnal valves, 600 cec nds after scram, is found to be sufficient to prevent en*xf in*w the region where condensaticn instability may occur.
Cne additicnal cption which has not been analyzed is the use cf the tt'+ine bypass system. Cpening the :EP/ and dtnping de vessel enercy to the condenser will depressurice the core wiecut increasing the pool heatup rate which is characteristic of the use of additicnal safet41 relief valves. This methof would cbvicusly result in satisfactcrf pcd te.paratures. Ac ss to the ccndenser will be readily avain'ble in alrcot every situaticn.
"he analysis presented here represents a ver/ unlikely set cf events which have a Icw prcbcbility of occurren ;
hc.mver, emn if this SCF"I transient shculd occur, dere is sufficient conservatism in the Bl~:'..ns Ferrf ;uclear Plant desicn to premnt concurrent high mass flux and high *wrc.s terperatures by several alternate means..: is ccncluded, derefore, dat de rrshead discharge cev.ms can be cperated in a stable environment and that no cperational restrictions are needed.
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Re fe rences 1
PETPAN - A Program for One Dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, bolure I Equations and Numericc, EPRI-CCM-5, December 1978.
2.
Besidual Decay Energy for Light Water Peactors for Long-Term Cooling, Branch Technical Position ASB 9-2, U.S.
Nuclear Regulatory Commission.
479
E!!CLOSL'Ei: 2 A.
9 tuck-Open Sa f ety Felief Valve Transients discusses the various a na lyse s pe rf ormed for a stuck open relie f valve event at the Browns Ferry tiuclear Plant.
These analyses have been performed using very conservative assumptions.
The results of these analyses show an adequate margin exists between the prodicted maximum suppression pool bulk temperature and the limit for stable con densation (less than 1500 F when the exit mass flux is greater than 40 lb /sec ftz) whon both trains of heat exchangers are available or wnen operator act ion is taken to cpen an additional relief salve.
Appropriate actions such as these already form a part of the plant operating procedure.
3.
Suppression Pool Temperature Monitoring System
- he suppression pool temperature mor.itoring systen consists of se ve ral instruments located in the tores and the lines which take sucticn from or discharge to the tcrus.
These devices are listed in the follcwing table and their location relative to SPV discharge positions is indicated in the follcwing sketch.
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Function In9trument I f.ocation Torus Temperature TI-64-55 Torus "emperature TI-64-SS R ER S uction TW-74-9 Heat exchanger inle Loop A R HP Suction TW-74-32 FMat exchanger inlet Loop B R HB Suction TW-74-21 Heat exchanger inlet Loop C R HR Suction TW-74-43 Heat exchanger inlet Loop D RHR Cooling Return TW-74-81 Heat exchanger outlet Loop A R !IP Cooling Return TW-74-82 Heat exchanger outlet Lool B PHR Cooling Return TW-74-83 Heat exchanger outlet Loop C R HR Cooling Return TW-74-84 Heat exchanger outlet Loop D
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BROWNS FERRY NUCLEnR PLANT UNITS 1-3 TORUS TEMPERATURE MONITURING SYSTEM i
f f
180 I
l B
L
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F C 3-C' S
O' TORUS WATER TEHP
'O G
E RHR SUCT10N V, ">
p RHR SUCTICN O
270 900 H
D ;,
q LJ L3
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A K
r3 A
C G
v N RHR SUCTION RhR SUCTION L3 C O 0
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E:;CLOSm 3 SIGNIFICANT CONSERVATISMS !h THE SORV ANALYSIS The analvr.ed case assenes a valve is opened above 120 F pool temperature with no heat exchangers in operation in one case.
As in all transients, the reactor is considered to be initially at a steady state power of 103 percent.
The suppression pool is assumed at its maximum technical specification terperature of 93 F and its minimum water volume.
122.5 percent of ASME related SRV capacities have been assured.
During the SRV transient, the pressure in the vessel would initially decrease due to the larger energy removal rate not immediately acconnodated by the reactor system. The turbine control valves automatical!y adjust during this time in an attempt to maintain th2 reactor st 'am dc=e pr: ssure. The vessel pressure partially recovers due to this a,
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dowever, to simplify the analysis, this transient is ignored nnd it is assured that the initial pressure in the vessel is maint.ined through this period.
The reactor system dumps steam to the pool at full power until the pool reaches 110 F at which time the reactor is scra==ed by operatcr action as required by the tachnical specifications.
This action eccurs approxinately '00 seconds into the transient. Ccnservatively.
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the main steam isolation valves are used in this simulation to I
isolute the core. The MSlV's are assumed to renain open until a core low water level signal is received. Minimization of this time i
and consequently the energy lost to the turbine is obtained by closing tha feedwater input (by losins offsite power) to the vessel at the time of scram and reinstating cooling water only after the MSIV's are closed. This operation maximizes the energy released to tne pool since the core is " bottled up" at a higher cecay heat power level. No energy is lost to the condenser by way of the turbine bypass syster since these valves are lef t arbitrarily clcsed throughout the sL=ulation.
Nor:al plant operating experience indicates that such isolation does not occur.
Without isolation, a significant quantity of steaa may be dumped to the condenser throughout the stuck open SRV transfeat, thereby further liti ting the increase in suppression pool water temperature.
Several indications of SRV opening are available to the control operator including load reduction, change in measured seca: flow, root compensating turbine bypass valve closure, rise in SRV discharge line temperature (recording and alara), and acoustic tonito-ing of SRV discharges. The above indications assure that the op rator will be i==ediately aware that a SRV has inadvertently opened, and that he can quicklj-pinpoint which SRV has opened, so that the proper actions cay be taken on a tirely basis. During ac tual plant operation, suppression pool cooling is initiated promptly upon SRV openinz.
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Cne-half of the ?,liR suppression pool cooling capacity has been assured to be inoperable.
Forty years of crud accumulation has been assumed on the RHR heat exchangers.
Lased en the above and the analyses presented, TVA believes the Browns Ferry design has been demcnstrated to be satisf ac tory and conserva tive,
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