ML19250C463
| ML19250C463 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 11/13/1979 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| LAC-6611, NUDOCS 7911260164 | |
| Download: ML19250C463 (53) | |
Text
3 DAIItYLAND POWEIt COOPEIRATIVE Ba Grona, OYaconaa 54601 November 13, 1979 In reply, please refer to LAC-6611 DOCKET NO. 50-409 Director of Nuclear Reactor Re n tation ATTN:
Mr. Dennis L.
Ziemann, Chief Operating Reactors Branch #2 Division of Operating Reactors U.
S.
Nuclear Regulatory Commission Washington, D.
C.
20555
SUBJECT:
DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR)
PROVISIONAL OPERATING LICENSE NO. DPR-45 ADDITIONAL INFORMATION REQUIRED FOR LRC STAFF GENERIC REPORT ON BOILING WATER REAC". ORS
REFERENCES:
(1)
NRC Letter, Ziemann to Linder, dated July 17, 1979 (2)
NRC Letter, Ziemann to Linder, dated July 27, 1979 (3)
NRC Letter, Ziemann to Linder, dated July 31, 1979 (4)
NRC Letter, LAC-6486, Linder to Ziemann, dated September 7, 1979 Gentlemen:
Our response to Attachment 3 of Reference (1) is hereby submitted.
I.
Do proceduma exist fcr loss of feebater and small break LOCA fcr the cases of:
(l) all AC pocer available, (2) loss of offsite power, and (3) loss of all AC pacer?
RESPONSE
A copy of the procedure for loss of feedwater is enclosed and identified as Alarm B13-1 (Bearing Oil Header Pressure (Lo)
R.F.P.
Trip).
This procedure is a part of the plant Operating Manual.
The scram procedure Section 3. 3.1 which is referenced in the RFP trip, is also from the plant Operating Mar.ual and is also enclosed.
The small break LOCA procedure is designated as Section 3.3.3, Primary System Leak, which is enclosed and is also a part of the plant Operating Manual.
Enclosed procedures for loss of feedwater and snall break LOCA are applicable to all three cases addressed in the question.
1395 019
,eo y j 1-9
Mr. Dennis L.
Zicmann, Chief LAC-6611 Operating Reactors Branch #2 November 13, 1979 II.
Discues the reactor water level measurement system.
In particut' -
1.
Provide a diagram shoving location of pressure tape used in measuring level.
The diagram should be dstailed enough to shou uhether the measurement is inside or out-eide the core ehroud.
RESPONSE
Enclosed are Figures 4. 53 and 4. 67, extracted from the La Crosse Boiling Water Reactor (LAC 3WR) E.feguards Report for Operating Authorization which together show locations of pressure taps used in measuring reactor water level.
The core support skirt is essentially an open bottom solid wall cylinder up to the core grid plate.
This core grid plate has 72 in-dividual holes matching the locations of 72 fuel elements.
Each fuel element is surrounded by a solid walled fuel element shroud.
The peripheral shroud is open at the top and sides.
Therefore, the water level in the core is only a valid neasurement when the water level is higher than the fuel element shrouds (elevation 664' 7.75").
2.
Describe the instrument piping arrangemente and types of transducere used.
RESPONSE
The instrument piping arrangement is shown on the enclosure labeled
" Sketch 10872-296".
The transducers are Model No. 613DM-MS2-0 manufactured by the Foxboro Company.
An enclosure,18-155, gives a description of the transducers used for water level.
3.
Which levels are monitored in the control room and hou are they indicated (i.e.,
recordere, metere)?
RESPONSE
Figure 2.10, enclosed, is a page from the plant Operating Manual and shows the type of monitor from each level transmitter.
All indicators shown on this figure are in the Control Room.
Figure 2.6, also from the plant Operating Manual, shows the type and location of all level indicators.
Indicators 50-42-812 and 50-36-803 and the level guage glass 50-42-803 indicate only locally. 1395 020
4 Mr. Dennis L.
Ziemann, Chief LAC-6611 Operating Reactors Branch #2 November 13, 1979 4.
Which measuremente provide signale for safety systems, which for control systems, which for other eyetems?
RESPONSE
A description of the functions performed by the signal from each of the transmitters is enclosed as Section 2.3.3, Reactor Vessel Water Level Instrumentation, a part of the plant Operating Manual, and is summarized in the table below.
TRANSMITTER 50-42-XXX Function 301 302 30 3 304 305 306 Lo Water Level Scram X
X X
Hi Water Level Scram X
X Lo Water Level Alarm X
Hi Water Level Alarm X
Start Emergency Core Spray Pumps X
X X
Start Emergency Power Diesel X
X Open Alternate Core Spray Valves X
X Isolate Certain Containment Valves X
X Feedwater Control X
Open Low Pressure Cu - Spray X
X S.
Describe the dynamic response of each of the level measurement ar.d indicating instrumente for conditione typical of a small break LOCA.
DPC RESPONSE:
~
All the level transmitters, except 50-24-304, are connected to the vessel at the same height for bcth the high and low legs, through b
s t
Mr. Dennis L.
Ziemann, Chief LAC-6611
.perating Reactors Branch #2 November'13, 1979 identical penetrations.
All transmitters and amplifiers are of the same type and would have the same response and accuracy.
Therefore, all water level transmitters, except 50-42-304 would have the same dynamic response to a small break LOCA.
Enclosed is Section 5, Effects of Transients on Water Level Indi-cation, from a Gulf United document SS-ll82, Evaluation of Reactor Vessel Water Level Indication for La Crosse Boiling Water Reactor dated March 5, 1974.
Also enclosed is a partial Section 5, LACBWR Loss of Coolant Accident, from a Gulf United Document SS-942, Technical Evaluation Adequacy of LACBWR Emergency Core Cooling System dated May 31, 1972.
6.
What are the level measurement uncertainties?
RESPONSE
An excerpt from ACNP-66548, Amendment to LACBWR Safeguards Report for Operating Authorization, numbered II-3-26, and -27 is enclosed describing water level instrument errors.
7.
What level difference is expected betacen core and measurement location for:
normal operatione, a.
b.
reactor shutdoun uith decay heat and uith recirculation pumps running, c.
Reactor shutdoun with decay heat and recirculation pumpe not running, and d.
moderate level transient ao for a enall break LOCA or stuck open SRV.
RESPONSE
For all conditions, when water level is above the fuel element shrouds, no level difference is expected between core and measurement location.
If there are any questions concerning this submittal, please contact us.
Very truly yours, DAIRYLAND POWER COOPERATIVE
.44 Frank Linder, General Manager FL: HAT:af Enclosures -
h o
3.
Check system for leakage.
m=
S $
4.
If system leakage cannot be quickly repaired, start the l3 )
standby reactor feedpump and secure the running reactor feedpump for maintenance.
5 7
5.
Contact instrument department to replace or repair level E
switch.
? c i oo
-o ;..
i ALARM B13-1 (BEARING OIL HEADER PRESSURE (LO)R.P.P. TRIP).
7 E
Possible Causes 9
1.
Hearing oil pressure has droir ca to 5 psi.
.u.
2.
System leakage has dropped reservoir level causing loss of pump suction.
Immediate Action.
1.
Acknowledge alarm.
2.
Start the standby reactor feedpump immediately to avoid E
reactor scram, o
~~
3.
If the feedpun.p cannot be started and loaded properly in j
time, proceed with full scram procedure (Vol.
I, Sec.
- 3. 3.1).
2, E
4.
Determine cause of pressure switch actuaticn.
n Eo U
?
ALARM B13-2 (l'EEDWATER AUXILIliRY OIL PU::PS AUTO-START)
" i E N Possible Causes.
E r-1.
Bearing oil pressure dropped to 10 psi.
3 2.
Running pump tripped on overloa;.
73 3.
Running pump output pressure setting reduced too low.
4.
Loss of pump suction due to low reservoir oil level.
\\_
5.
Pressure switch malfunction.
m C
g Immediate Action E
1.
Acknowledge alarm, e
t U
2.
Determine if lock nut has vibrated loose on pressure set et.,.
screw, causing purp output pressure to sag.
If so, restore outlet pressure to approximately 20 psi and tighten lock nut.
u O
3.
If pump tripped, determine cause.
=
=
g ~.'
3b If lube cil reservoir low, the a]arn "PEEDUATER BEARING 4.
I OIL TAMK LEVEL LOM" sl.ould have annunciated with resulting immediate action alrcad; taken.
CD E.
5.
Contact instrument department to replace or repair pressure
?,, h, switch.
til J. yap 1m 1395 023 W
me o
-s LACBWR Operating Manual Revised i
Volume I, Integrated Plant Operations October
~5 3.3 EMERGENCY PROCEDURES 3.3.1 Scram Procedure (Full or Partial) 3.3.1.1 Symptoms.
(1)
Alarm D4-1 (Reactor Partial Scram) or (2)
Ala rm D4-2 (Reactor All Rod Scram)
(3)
Control Rods Scram to their " FULL-IN" position.
o s
I (4)
Reactor power is decreasing.
j (5)
Alarm on Panel D indicating in " Red" is cause of scram.
2 N j
3.3.1.2 Automatic Actions.
[
]
(1)
All rods are inserted and power is decreasing for " FULL SCRAM.
U
[
(a)
Rods 1 thrcugh 13 are inserted and power is decreasing i
NN for " PARTIAL SCRAM."
y,1 (2)
Forced Circulation Punps reduce to 80% speed (approximately 6
3 0I 800 rpm) if above that speed at time of scram.
(3)
Station load transfers to the Reserve Auxiliary Transformer.
E kl o 4 3.3.1.3 Immediate Action.
_t (1)
Verify that the specified automatic actions occurred.
j 4 (2)
Insure that the Reactor Feedwater System is maintaining i }
normal water level.
s 1
(3)
If a low reactor water level of -12 inches is reached, h
insura at least ene High Pressure Core Spray Pump is running.
g (4)
When generatcr load reaches 2 to 3 MWE, OPEN 25NBl and 3
switch the Regulux control switch to " MANUAL."
8 M (5)
Trip turbine.
o h.
(6)
Adjust the Main Steam Bypass Valve Controller to maintain 3 Q pressure, if necessary.
I k (7)
Turn Log N Channels 3 and 4 key bypass switches to "NO RM AL. "
~
g (8)
Turn on High Voltage to Source Range Channels 1 and 2, when neutron level decreases to 10-10 ampe on Channels 3 and 4.
7 5
(9)
Down scale Wide Range Channels 5 and 6 Range Switches to monitor reactor power level.
v B o m
3.3.1.4 Subsequent Action.
7 k E (1)
Notify the dinpitcher of scram.
o E d (2)
Determine cause of scram and time required before rods can
^
be withdrawn.
I f great or than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, drive remaining rods in using the All Rods Inser*,
6" o Partial Fcram has occurred.
t 116 0'D rp[ryrunNfd $]_$
blos i395 it'4 ia
. y....
LACBWR Operating Manual Revised Volume I, Integrated Plant Operations October 1975 n
s 3.3.3 Primary System Leak 3.3.3.1 Primary System Leak Detection.
The Primary System, for this procedure, is defined as the Forced Circulation System, Primary Purification System, Primary Steam Lines to Shutdown condenser, Relief Valves and Main Steam to the Main Steam Rotoport Valve.
The Feedwater Header and piping to the Feedwater Check Valve in the Containment Building is also considered e
8 as the Primary System.
A leak from the steam or water phase within the Containment Building
_j J or lower cavity will be detected by one of the following:
- (1 a
(1)
Humidity detecting and recording equipment, detecting g
}
changes in humidity.
u t
(2)
Continuous activity air monitors:
E g
(a)
Containment Building Monitor 3 q (b)
Mobile Monitor N
(c)
Cavity Monitor Y
(d)
FC Cubicles Monitor
.e o
(3)
Visual inspection.
e An alarm from the Hwmidity Detecting System or the Continuous Air j y Monitor shall be investigated immediately.
An inspection of the g
(i Containment Building will be made in an attempt to locate the source of the leak.
Air samples shall be taken to deternine whether the g
I leak is a steam or water phase leak by isotopic identification.
u l
The change in the humidity or concentration of the radioactivity g
i will be used to estimate the magnitude of the leak and also source.
a g
. A known leak in the Reactor Lower Cavity will be menitored every y
n, shift to estimate the rate of increase of the leak.
The actior.
o
[f taken will be determined by the size of the leak and the rate of g Q increase.
For a large leak, follow procedure 3.3.3.2.
If a leak jo g is small and the rate of increase can be determined, the activity
~
discharge from the stack will determine the course of action.
gr E
a_
Dk*v;D'9'T'6I D
i 2
. ca. 3L 1395 025
-12",
the Main Condenser E '
can be used as a heat sink.
^'
1395 027 L 116 3-170
LACBWR Operating Manual Revised Volume I,
Integrated Plant Operations July 1976 (b)
The " Overhead Storage Tank Level Lo-Lo" alarm annunciates and reac tor water level is being maintained (supplied) from the Overhead Storage Tank.
NOTE:
If unable to maintain OHST level above the Lo-Lo level alarm and there is the possibility of needing A
Building Spray System to decrease building pressure, h
then "OPEN" HPSW to Core Spray Pumps" supply valve o
j R by actuating acontrol switch on Benchboard E and cross-connect "HPSW to Demine ralized Wate r" system
{
by opening manual valves located in Turbine Building above 1B air compressor.
_g k.
9 (c)
All personnel are evacuating the site and the
,y
- [
primary leak is continuing; the following " NOTE" may be used:
NOTE:
If ACS is required and Primet, System pressure is too
[ h high
(> 50 PSIG), manually "OPEN" both " Reactor g f, Emergency Flooding Vent" valves using keyswitches on 3.0 Vertical Panel "D" to rapidly depressurize the vessel.
3]
CAUTION:
When Reactor Emergency Flooding Vent Valve 1A " OPENS,"
e the demineralized water header isolation to the
$ C shutdown condenser " CLOSES" to ensure that full a E demineralized water flow will be making up to the OHST.
j (9)
Power feeds to the building must be tripped.
This is k
done by placing the " REACTOR BUILDING MCC FEEDS - MCA TRIP" E
control switch in the " TRIP" position.
Start 1 B cn' pump.
d q
E
\\
(a)
Place in " PULL-OUT" control switches for the
[.
following:
\\
-Shield Cooling Pump 1B
$ b
-CRD Ef fluent Pump 1B j
-Fuel Storage Well Pump 1B o
<o
-1A and 1B Seal Injection Pumps
, n
.j s (b)
Ensure " Control Rod Charging" pump switch on BBD "D" l
O is turned "OFF."
5 (c)
At 480-V Switchgear 1B, " TRIP" Reactor Building Crane
?.
Feeder Breaker.
(10)
If the Alternate Core Spray System is needed and did not
$ g, start automatically, it must be started as follows:
u (a)
Place either " Diesel H.P.
Service Water Pump 1A" or h
" Diesel H.P.
Servicc Water Pump 1B" control switch in the 2
- Control Room to "RUN."
(b)
Get keys from Shift Super"isor and place " Alternate Core Spray AC talve" and " Alternate Core Spray DC Valve" keyswitches in the "OPEN" position.
) ') (;'
tj g
,r L-116 3-171
e LACBWR Operating Manual Revised Volume I, Integrated Plant Operations June 1976 (11)
If the water level cannot be maintained in the reactor with the one Alternate Core Spray Pump running, start the second diesel pump.
NOTE:
If core level is being maintained without high pressure service water being admitted, secure the engine driven service water pumps.
If level is being maintained by intermittent cycling of makeup valves, secure one R
engine driven pump and allow other to operate.
E o o q (12)
If all personnel are evacuating or when the water level y
in the Containment Building reaches 60 inches on the " Containment Vessel Liquid Level" indicator, "OPEN" the " Reactor Emergency 4
-j Flooding Vent Valves" to equalize water level in the building and g
reactor vessel.
NOTE:
If Contninment Vessel Internal Pressure increases to b .
52 PSIG, "OPEN" the " Building Spray Valve" located behind the Main Contr 1 Panel as necessary to hold I 4 the pressure below 52 PSIG.
t j d (13)
When the water level reaches 50 inches on the " Reactor i
Vessel Wide Range Level" indicator or 330 inches on the " Containment
- 7, Vessel Liquid Level" indicator, the Alternate Core Spray System 2 O should be stopped.
The water level should not be increased much 5 $
beyond this point because it will reduce the volume available in
' -)
the containment vessel for pressure buildup.
8 (14)
If the Feedwater Pump trips or if all personnel are evacuating the Control Room, "CLOSE" the Reactor Feedwater Flow R
Control Valve.
5u (15)
The Shif t Supervisor will evacuate the remaining g f personnel unless, consistent with personnel safety, it is feasible
\\ for at least two (2) men to remain in the Control Room to monitor
[
plant conditions.
o
,7 (16)
Notify the following plant personnel:
~
? Q l5 3 (a)
Operations Supervisor (b)
Assistant Superintendent 3
(c)
Plant Superintendent g
(d)
Health and Safety Supervisor
?
e j S (17)
For long term cooling with the Containment Building g
flooded to 330 inches on " Containment Vessel Liquid Level," CCW will be continued to the Containment Building for heat removal.
u E'
E.
NOTE:
Containment Building insulation can be removed to aid E '
in heat removal.
i'"
n70 Ul!
t J
=
t 116 3-172
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=
REACTOR INTERNAL COMPONENT ELEVATIONS FIG,4.53
U lf I
c, s
peripheral shroud
(
3
/j
\\
separator. let in x
l>/
x
/
l 6
i I
l thermal shield fuel element shroud i
l
,j L
K core grid plate I,
2 4(trt este wr +
plenum separator plate /
reactor vessel 2(($[$3, I395 05I SCHEMATIC OF THERMAL SHIELD-iO-VESSEL ANNtJLUS FIG. 4.67
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- gSUPPLY, 4
y \\< s I b I I E4976 l'ig. 1 The Type 613DA! D:fferential Pressure Trans-on the diaphragm-capsule, which is rigidly con-mitter mea su res diff erent ui pressures in the nected to the force bar ills by the "C" ticxure r naes of 0-20 to 0-850 inch s of water. The i9i. The Eleilcy diaphragm e 10 s acts as a seal di!!erential pressure is deteete d by the trans-and as a fulcrum for the force bar. The force mitter, and co:werted to a prop utto ul 10-50 ma bar transmits a force, which is exactly propor-d-c s!gnal which as t ran smit t ed ) a remo:e tions! ta the differential pressure on the dia-Electronic Consotrol Receiver. phrag'n-capsule. through the flexure i12) to the range rod i!I causing the range rod to pivot Refer to Fig. 1. about point #2). Pressures are applied to oppmite sides of the Any movement of the range rod is detected by silicone-filled d bph ragm-capsule e3 through the armature s3 of the differential-trar.sformer the high and low pressure connections. Any detector #4 s. If. for instance. there is an in-di!Terence between there pressures exerts a force c rease in pressure on the high pressure side of ^ D M ~ i395 0.53 a The Foxboro Compmy Printed in U S A. Foxboro. Mass. l' S A FO.MM 5.000--12/61
13-135 Pa;:e 2 the diaphragm-capsule, the armature moves output is exactly proportional to the diacrential toward one secondary winding (direction shown pressure. In operation the movement of the range by arrown of the detector, thus strengthening the rod is continuously adjusting the detector e 4) to inductive coupling and increasing the secondary maintain a condition of force balance between voltage. IIence the 10-50 ma d-c output signal the forces exerted by the feedback motor and by from the oscillator-amp!!.*.cr will increase. This the force bar. output signal is fed to the feedback motor (55,in Two vent screws (61 and three drain plugs 871 series with the receiver. The force exerted by the are provided for removing entrapped att or feedback motor is exactly proportional to the liquids. force applied to the range rod (16 by the force bar (111. Since the force exerted by the force bar A stabilizer is provided in tha topworks to damp is exactly proportional to the dicerential pressure, out vibration which may occur during ce!! oper-the current in the feedback mc, tor and to the ation. Specifications Measurement Range: 0-20 to 0-205 inches of water 10-500 to 0-5100 mm of wateri, or 0-200 to 0-350 inches of water (0-5000 to 0-21.000 mm ef wawri Output Range: 10-50 ma d-c into 600 ohms 10 percent Accuracy: 0-20 to 0-205 mehes of water,20.5 percent of full span; 0-200 to 0-450 inches of water,10.5 percent of full span; 0-450 to 0-750 inches of water, 21 percent of full span; 0-750 to 0-850 inches of water, el percent of f ull span up to E5 per-cent of span, and -2 percent at 100 perc(nt of span Maximum Working Pressure: 1500 psi Working Temperature Limits for Cell Body: 40 F to -250 F Ambient Temperature Limits for Amplifier: 20 F to -180 F Power Supply: 65 volts d-c. 2.5 percent Process Connection: Flange connectors, either 3 cinch or 12-inch NPT, or 32-inch Sched-ule 80 wc! ding neck I:1cetrical Connections: Tapped for tz-inch conduit Weight: 28 pounds Power Consumption: 5 va d-c max n.um Hody: Car tmn steel, Type 316. stainless steel, or Monr1 metal Topworks Cover: Cast aluminum, weatherproof, gray finish Diaphragm-capsule: Type 316 stainless steel, or Monel metal r i395 034 The Foxboro Company Prmied... IJ S A Foxbaro, M m. t! S A M@d
LACBWR Operating lbnual, Volume it, Reactor Process Systens Revised October 1%'8 s 9-l ( _, - - ~ [ 'N'Rf" Tar <3 s e f3 CJ2 l
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7/rArp, j,9 A', f,,, Wr/o r-o Level Transm.itter Level Indicators / Recorders Norrnal Readi3 (with ranges) +30" -g 5c" 50-42-331 LR 50-42-801 0" } re x -3c" ?% _ Feed Purp ,y 6 -- Control ] gc ,39,,' f 30' q 9 go
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LACBWR Operating Manual Revised Volume II, Reactor Process Systems July 1978 9 %^ e E ) 1 2 b< 2.3.3 Reactor Vessel Water Level Instrumentation \\ b k' Reactor water level instrumentation provides indication of reaucor kq water level, automatic level control, and high-and low-level scram s E signals. Level probes are also provided to detect the actual water g g surface pattern during power operation. A Bi-Color level gauge is 3 D. provided to show reactor water level and may be used as reference N to calibrate other level instrumentation. e' A 2.3.3.1 Water Level Transmitters. Six reactor water level trans-9 mitters are installed to measure the water level in the pressure 0 vessel. Three transmitters, 50-42-302, 50-42-303, and 50-42-306 provide signals to the Reactor Safety Systems Water Level Channels 8 No. 1, N o. 2, and No. 3 (Volume IV, Section 5, Safety System). $ ) In addition, transmitter 50-42-303 supplies a signal for the g REACTOR' VESSEL WATER LEVEL indicator in the Control Room located on Panel E. A third transmitter, 50-42-301, supplies a signal to og( the Control Room for recorded level information on REACTOR WATER g e LEVEL recorder which has a range of -30 inches to +30 inches and also provides a signal to the Reactor Level Controller ( Re f. : cy y Volume IV, Section 8, Reac tor Water Level Control). The reactor \\ level controller is part of the 3-element level control system that varies the feedwater pump speed, maintaining water level auto-e o g g matically in t he pressure vessel. The fourth level transmitter, c 50-42-304c provides a signal to an additional REACTOR VESSEL HIGH LEVEL indicator which has a range of 0-150 inches. This indicator e is particularly useful during reactor flooding and maintenance when } the reactor is sh_.down. The fif th level tras;mitter, 50-42-305, 5 g provides a signal to REACTOR VESSEL WIDE RANGE LEVEL indicator on E j Panel D-1. Its range is 0-250 inches which provides wider indica-a tion in event of loss of primary coolant. o e n g g The sixth level transmitter, 50-42-306, for reactor water level e Channel No. 3 provides a signal to the start circuits of the core ag ' o spray pumps at 12", a 'ow water level scram signal at -12", and a signal to water level 2ndicator on Panel E. It has a range of -30" to +30" and provides a redundant indication of reactor water level. '"o 2-5 1395 037
LACBWR Operating Manual Revised Volume II, Reactor Process Systems July 1976 The REACIOR WATER LEVEL CHANNELS NOS. 1 and 2 and level recorder power supplies are turned on at the chassis. See Figure 2.3, showing i'dicated inches alarm and scram points vs. elevation in reactor vest 31. 2.3.3.2 Reactor U r~el I ~te. Gauge. The reactor vessel level R gauge is a 3-sec'TBn Duraport, 3000 PSIG gauge with illuminator ~ A and variview wiu.
- yle vision hood.
The hi-intensity illuminator j C has t.lgh br.llar reflector type lamps that direct a beam of light N through a verticuily divided red and green color screen and strip lens into the ports on the rear of the gauge providing color 3 discrimination between water and steam. The variview hood provides maximum viewing angles by use of combination frosted glass and [ j(ymirrors. The ' auge allows a 60-inch span of vision. See Figures 2.4 and 2.5 for indicated level vs. actual water level in the reactor. E a l 2.3.3.3 Liquid Level Probes. Liquid level probes are provided to g y check the water steam interface level in the reactor. Six sample probes are capable of being installed in the pressure vessel beginning 3 2 feet above the core at the 665'8", 666'8", 668'0", 669'4", 670'8", 3 and 672'0" levels. (Sco Figure 2.6.) The water sample from these elevations is directed to a cooler and flowmeter. The change in condensate rate as reactor water is raised or lowered will indicate e j A the water level in the pressure vessel. The use of the liquid level probe system will provide water level information to compare with ' water level transmitters and level gauge readings. This information will provide curves that will show actual water level vs. indicated ? level, for various reactor powers, thus showing true water level from J level instrumentation readings. [ The probes are fabricated of S.S. 304 1/4-inch tubing with 0.035-inch wall thickness. The six probes may extend in length from 2 feet to v 8 feet 4 inches above the core. Guide tubes are provided for support [I g of the probes and extend each probe end. The end of each probe is h[ 0.040 inches in diameter and 2.00 fitted with an end plug of the sample material. The sample hole is [ inches from the tip of the end plug. o 6 1 The upper end of the each probe penetrates the 20-inch access flange C' on the reactor head. The penetrations are fabricated with 1/2-inch j$ 0 S.S. 316 Schedule 80 pipe penetrating the 20-inch flange welded to sb the flange at both ends, with reducere and tube fittings. The sample a lines are fitted with double shutoff valves prior to entering a 7 header to the sample cooler and flowmeter. j The condensate measurement system consists of a cooler, flowmeter, E pressure gauges, and temperature indicators. The sample flows from 3 $ the probe to the sample header past a pressure gauge, through a u di cooler and flow indicator. The shell side of the cooler is supplied g with component cooling water. a a 1395 038 L - 116 2-6
LACBWR Operating Manual Revised Volume II, Reactor Process Systems July 1976 2.3.3.4 Valves. The shutoff valves associated with the reactor level piping are globe valves designed for 1500 PSIG at 10500F working pressure and temperature. The valves are constructed of 316 S.S. integral seat-back seating type with socket weld joints. The valves have been hydrostatically tested at 1 times design working pressure. .A 2.3.3.5 Piping. All piping associated with the reactor level y q piping conform with ASTM specification for H376 Type TP-304 Schedule 80 seamless pipe. The piping is designed for working pressure and temperature of 1450 PSIG and 650oF. The piping and i valves have been hydrostatically tested at 1 times design working g pressure. a 2.3.3.6 Stand-Pipe. The " Stand-Pipe" provides the variable leg 5 for the reactor liquid level transmitters. The stand-pita is ,? constructed of 304 S.S. 2.375 OD and 1.939 ID, with a length of j.I 20 feet. The stand-pipe is attached to the reactor vessel with the upper connection at the 677'l.75" elevation and the bottom , connection at the 657'0" elevation. The liquid level transmitter I ('} 2 low pressure taps are connected to the stand-pipe at the 663'11" elevation giving a 13'2.75" active length. The stand-pipe is r covered with a minimum of 3-inch insulation. The insulation main- . s j A tains the water in the stand-pipe at near reactor temperature reducing error of the level instrumentation caused by temperature. Q -? 3.3.3.7 Condensate Reservoir. Four condensate reservoirs are provided, one for each of the four transmitters (50-42-301, 50-42-302, 50-42-303, and 50-42-306). The reservoirs are located at ) he top of the transmitter reference legs and two of them connect c 3 l to interconnecting piping between the reactor and the top of the stand-pipe at the 677'l.75" elevation, whereas, the other two are 2 connected directly to reactor vessel penetrations. The reservoirs remain uninsulated to allow steam to condense keeping the reservoir s j % full, maintaing a constant level in the reference leg of the transmitters. The condensate reservoirs are fabricated of 900 SS A R
- S.S. T-304 inch pipe, 16 inches long, with appropriate pene-
,,j k trations. One 1-inch bottom penetration connected to the trans-I C mitter high pressure side, and a 1-inch penetration connected to the reactor vessel. (See Figure 2.7). ,m 3 2.4 OPERATING PROCEDURES 5 2 2.4.1 Placing the Reactor Vessel Level Instrumentation In-Service e U Prior to Reactor Startup. o 3 $ i " 1375 039 r L - 116 2-7
LACBWR Operating Manual Revised 3 Volume II, Reactor Process Systems October 1973 2.4.1.1 Prerequisite. 1. The Reactor Vessel Valve Checklist has been completed. 2.4.1.2 Procedure. e 1. Close transmitters drain valves, o 2. Open transmitter equali zer valves. ]> 3. Open transmitter sensing line valves, Q 4. Back-off vent screws on each side of diaphragm and bleed g air from chamber. 8 5. Close transmitter equalizer valves. L 3 6. Check indication against level gauge indication. If the E S level transmitter indication does not acree, back-fill the referenc column by adding demineralized water (using hose and demin. supply) qg to the reference line through the drain valve until the indicated level ceases to change (reference leg is full). Close the drain e {ps [" 3 0 valve (disconnect hose), again vent air from chamber, and assure -3 that indication agrees with reactor water level gauge. If it appears tnat the transmitter or readout is out of calibration, noti.4 s g N the instrument department. ET I 2.4.2 Placing the Level Gauge Glass in Service with the Reactor g s Pressurized u a p 2.4.2.1 Procedure ()y 1. Open the drain valve and slightly open the stcam inlet vals o N allowing condensate to flew past each port for 15 minutes. If the Ball check valve closes, simply close the steam inlet valve and a( again adjust slightly open. o O m 4$ 2. Close drain valve and allow the gauge to fill with o g condensate and pressurize. g ? 3. Gradually open the steam and watcr inlet valves antil fully j j open, and water level stabilizes within the gauge. o 0 0 4. Turn a-c lighting power on at lighting cabinet, c m ao C-1395 040 L 11S 2-8
LACBWR Operating Manual Revised Volume II, Reactor Process Systems August 1976 f Reactor Vessel Water Temperature 670' 6" au
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- REACTOR VESSEL WATER LEVEL - TEMPERATURE CORRECTION FIGURE 2.3
/ (Standpipe Heaters In Operation) W J
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LA'CBWII Operating Manual Revised Volume II, Reactor Process Systems June 1976 500 F 400 F 300 F 200 F 150 F
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1 i j 673' 0" l i ~ j ; O 10 20 3,0 ' 40 50 60 70 80 90 100 110 120 130 140 150 3 ,/ indicated level, inches alternate core spray 1ine removed or drained i trans 50 42-304 e j indicate 50-43-809 a a M REACTOR VESSEL HIGH WATER LEVEL - TEMPERATURE CORRECTION FIG. 2,5 t -;L I i mS oc m,,,
Reviscd 'I'ACINil Opera ting F.anual Volume II, Reactor Proce:,n Sy:; t ems July 1976 vent connection 1/2" A N N E C 4" pipe (900 SS A) e o s x 7e \\' a. >[ 4 f ^ s N c-2 3 ~ s s u f Nk C. 2 6 6 G 'l e_ socket weid % M coupling (typ. 3 pl.) C / 1" U _~ j , slope l>1ck 'I ^ to reactar { j ,I } I t A c A U =2 $ ^If'
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2,, m s 0 C material - stainless steel (T 304) 1 N con,truction - wc!ded E insulation - svare v g ? eo o g-c= D 1" ']~' cr -- to DP cell E D t S ?. = c.EQ N O"'E : Condensate Pot For Ib act or I,o ve 1 Traru, nit t er lia s No Ve n t. Co n n e e t i o n. 1395 044 L - 116 FIG. 2.7 CONDFNSATE POT FOR,UACIOR I!OUlD ll.VF L
3 Port Elevation J 1 670'-2.5" 6-0 2 669'-11.75" -O 3 669'-9.0" = O 4 669'-4.25" O 5 669'-3.5" O -O 6 669'-0.75" =- 7 668'-10.0" O 8 668'-5.5" O 9 668'-2.75" 0 10 668'-0.0" O 11 667'-9-1/4" O 12 667'-6.5" 0 13 667'-3.75" -O 14 667'-1.0" 0 15 666'-8.5" O 16 666'-5.75" 0 17 666'-3.0" 0 18 666'-0.25" O Nominal equivalent solio w tu lev I 666'-0" 19 665'-9.5" O 20 665'-6.75" 0 21 665'-4.0" ( 0 (Top of core 663'-8") 1395 045 REACTOR VESSEL LEVEL GAUGE FIG. 2.8
LACBWR Operating Manual Revised Volume II, Reactor Process Systems October 1978 REACTOR VESSEL WATER LEVEL CORRELATION (All Values Based on 5770F No Voids) RULES ELEVATION HARDWARE Level Transmi'- ech. Feet Abcvc k'ater Level Fater Level N SAR 50-42-303 Specs. Mean Sea Level Gauge Probec (Narrow Rang 8 $ 672' 0" No. 6 _g 670' 8" No. 5 S j Containment 5
- /
Building Limit 670' 7.19" (SAR 4.6.9.1 9% 670' 6" +30" U 670' 2.5" k 669' 7" Hi Level Scr 0 % Q. 669' 4" g g,g No. 4 4 >cL S o Ig qv $ s. q 668' 2" ~ k es u Normal Operat. 40 668' 0" 3b 0" $ D >c } Level i 2 (SAR Fig. 4.1) $3 667' \\" - c4 3 Lo Level Scr E ?US I4 a9 nds e r. 666' 5" E g n b oU } {= 665' 8" (gy No. 1 E NoN k 065' 6" $ o ~j og -30" c 8o % 665' 4" m S r-I D o n Top of Core r, Laterial Support c 664' 7-3/4" m Structure } (SAR Fig. 4.1) 5 m E $ Top of Core "b (LACBWR Operat. o 663' 8" ? u Manual Vol. II, c) g Figure 2.6) b $ FIGURE 2. c-1395 046 m
5. EFFECTS OF TRANSIENTS ON WATER LEVEL INDICATION A number of transients have been analyzed for the LACBWR system by means of the RELAP-3 Code" in order to investigate the response of the water levelindicator to changes in the level in the reactor vessel and to assess the adequacy of the levelindicator set points to safeguard the plant components. In selecting the transients to be investigated, previous analyses were examined to determine which incidents required the intervention of engineered safeguards systems triggered by water level indication for the amelioration of the consequences of the incident. Among antici-pated transients investigated recently" and summarized in Table 5.1, only the Loss of Feedwater Flow, Incident No.16, depended on a water level signal for initiation of the safety system. Another incident where water levelindication is of the utmost importance is the loss-of-coolant accident, analyzed in another report;" there the initiation of the high and low pressure core spray systems depend on a low water level signal. These two incidents, the Loss of Feedwater Flow and the Loss of Coolant are therefore discussed here in detail. In addition, in order to obtain a better insight into the response of the level indicator to rapid changes in pressure, power and flow transi-ents, the following transients were also studicd: 1. Increase in Pressure, patterned after the Generator Loss of Load, Item 1, Table 5.1. 2. Decrease in Pressure, patterned after the Initial Pressure Regulator Failure, Item 11 of Table 5.1. 3. Loss of Recirculation Flow, patterned after Item 13 of Table 5.1. 1395 047 5-1
p TABLE 5.1 - ANTICIPATED TRANSIENTS Time of Assumed
- Signal, No.
Incident Scram Signal seconds 1 Generator Loss of Load High power 0.2 2 Turbine Trip High power 0.2 3 Initial Pressure Regulator Failure High power 0.2 (Increasing Pressure) 4 Main Steam Line Isolation Valve High power 2.6 Clortre 5 Loss of Feedwater Heater None 6 Shutdown Cooling Malfunction High power 1 7 Isolated Loop Startup (Cold) None 8 Uncontrolled Rod Withdrawal None at Power 9 Uncontrolled Rod Withdrawal None at Startup 10 Loss of Feedwater Flow Low water level 14 11 Initial Pressure Regulator Failure Low pressure 19 (Decreasing Pressure) 12 Opening of the Turbine Bypass Valve Low pressure 42 13 Loss of Recirculation Flow None 14 Isolated, Loop Startup (Hot) None 15 Recirculation Flow Control Failure None (Increasing Flow) 16 Increase in Feedwater Flow None e 1395 048 5-2
4. Increase in Feedwater Flow, patterned after Item 16 of Table 5.1. 5. Increase in Power. Simple up and down ramps vere used to represent this transient. In the present analysis some of the conditions of the original analysis have been altered slightly in order to give more conservative l results; for example, the action of relief valves in limiting reactor l pressure has been ignored in this analysis, in order to obtain a more i i pronounced effect of pressure increase. The results of the transient analyses of the levelindication j system lead to the following conclusions: 1. Due to pressure perturbations occurring during the initial stages of a transient, the water level indicator instrumen-tation, principally because of the small amount of water i present in the standpipe, reacts more rapidly to changes j in water level than the actual water level in the reactor vessel. This means that the setpoints of the safety system are reached wellin advance of actual danger levels, lead-ing to increased safety. 2. In those incidents where the activation of engineered safe-i guards systems is dependent on signals given by water j levelindicators, the present safety system setpoints are such that the plant is safeguarded from unacceptable f damage. 5.1 COMPUTER MODEL RELAP-3" calculates flows, mass and energy inventories, pres-sures, temperatures and qualitics, and of particular interest to this study, mixture levels along with variables associated with reactor power, reactor heat transfer, or control systems. In the RELAP repre-sentation of the LACBWR ( Fig. 5.1), the reactor was divided into 7 dis-crete volumes, which in addition to an explicit representation of the standpipe, includes the core, the steam dome and steam separator (one volume), the downcomer, turbine, reactor inlet plenum, and the recircu-lation loop. In RELAP the volumes are connected by junctions repre-senting flow res% stances such as pipe friction and inertia, or pumps, leaks and fills. The standpipe is connected at its upper end to the S -3 1395 049 .W
1 4 TURBINE 0 11' TO 4* 3* 3 5' CONDENSER O G O n n STEAM DOME AND SEPARATOR 7 677.18' G77.18' d e 2-G- 5 E O O O n b m b 5 2
- s 8
663.67' CORE O 7' 1' a 8' n a PUM .M '6 INLET L___J PLENUM 9' 6 G57.00' ~ R ECIRC. SYSTEM G56./5' G56.75' 10' G54.00' O 647.75' G49.00' FEEDWATER 652.92* 626.00* 1r 9 it it 1r 9 1r p u 1 volume --*-- Junction 1* Fig. 5.1 - RELAP 3 Representation Of LACBWR Plant 5-4 1395 050 /
3 steam dome and at the bottom to the downcomer through 2-inch pipes about 10 feet long. The model accounts for the friction and inertia in the pipes. To simulate condensation of the steam above the water level in the standpipe, an additional junction (No.11) from which steam can be extracted has been added to the standpipe. The model did not include neutron kinetics of the core. Rather, power history was an input obtained from previous transient analyses.18 The input parameters associated with the volumes and junctions are described in Tables 5.2 and 5.3. Init!al. pressures, temperatures and flows, representing full power operation, r.re also given in the tables. The output from the RELAP cale (an example is shown on page A-14 of the Appendix) gives the average pressure, temperature, enthalpy, and quality of each volume at selected time intervals; also the bubble mass, liquid mass, total mass, and mixture level of each volume and the flow rates in and out of the volume. Thus, complete information is obtained from the computer runs on water density and void swelling effects, as well as voided unvoided, and indicated water levels. 5.2 INITIAL CONDITIONS The initial conditions for all transients investigated were taken to be those for steady-state operation at 100?o power of 165 MW. Basic t reactor parameters are given in Table 5.4. The water in the standpipe was taken to be at the saturation temperature of 577* F with the space above the standpipe water level filled with saturated steam. A steady state run was made with the RELAP code to assure a balanced system for the initial conditions of each transient.- This steady state run re-sulted in a voided water icvel in the reactor vessel of 670.42 feet [ (corresponding to an unvoided level of 667.40 at 577 F), and an indi-cated level of 667.85 in the standpipe. These levels were then taken to be the normal levels (at time zero) and served as initial conditions for the transient analyses. The RELAP input for the steady state condition as printed out by the computer program is given in the Appendix. 5.3 LOSS OF FEEDWATER FLOW As indicated in Table 5.1, Item 10, one of the anticipated transients where a water level scram signal was taken into account in a previous analysis (Transients Without Scram 28) is the loss of feedwater flow. In 5-' 1395 051
TABLE 5.2 - VOLUME GEOMETRY AND INITIAL CONDITIONS Temperature, *F
- Volume, Height, Elevation,
- Pressure, or Average Quality 3
No. It ft it psia in Volume 1 Inlet plenum 2S 1.9 9.00 647.75 1309.93 505 0 2 Core 92.7 6.92 656.75 1306.11 3 Steam dome and separator 997.3 21.2 663.67 4 Turbine 100.3 10.0 677.18 5 Downcomer 324.0 13.5 652.92 1301.31 C 6 Recirculation system 404.6 28.0 626.00 1320.49 562.7 so W 7 Standpipe 413.2 20.2 657.00 1300.34 0.04692 o tn N
- At bottom of. volume.
c.n
e TABLE 5.3 - JUNCTION GEOMETRY AND INITIAL CONDITIONS Initial Friction Volume
- Flow, Coefficient,
- Inertia, No.
Connected Elevation Ib/sec sec /ft ft-8 2 3 1 1-2 656.75 2812.1 0.0000199 0.432 2 2-3 663.67 2810.1 0.00001574 0.417 3 3-4 677.18 170.7 0.000?,884 180.4 4 4-0 677.18 170.0 0.06052 0 5 3-7 677.18 0.797 11.82 1514 6 3-5 663.67 2658.3 0.0000107 0.43 7 7-5 657.00 0 11.25 1590 8 5-6 654.00 2645.6 0.00009296 13.7 9 6-1 649.00 2813.5 0.00005775 13.2 10 0-6 626.00 170.5 0.06221 0 11 7-0 677.18 0.770 3438. 0 1395 053 5-7
TABLE 5.4 -INITIAL CONDITIONS FOR TRANSIENTS, OPERATION AT 100% POWER Reactor Power MW 165 g Reactor Pressure, psia 1300 ^ Steam flow, Ib/sec 170 Core Flow. !b/sec 2814 Recirculation flow, Ib/sec 2644 Condensation flow through star.dpipe, ib/sec 0.77 Temperature of water in standpipe, *F 577 Level of water in standpipe, it 667.85 Level of steam water mixture in reactor vessel (voided), it 670.42 Feedwater enthalpy, Btu /lb 257 Average quality in core 0.0318 Average quality in steam separator region above core 0.06 1395 054 5-8
the previous analysis"it was assumed that the feedwater pumps trip and the feedwater flow decays to zero in 2 seconds. Based on mass balances of water in the reactor, a low water level scram point (-12 inch) was calculated to occur in 14 seconds. The resul'.ing scram signal would cause the main steam isolation valve to close, the shutdown condenser to be started, and high pressure core spray to be initiated, arresting the drop in water level before damage to the core could occur. As the original analysis was performed by means of the REPTA code," where the water level indication system was not modeled specifically, an analysis of this transient has been performed as part of the present study by means of the RELAP model described in Section 5.1. Specifically, it was desired to verify the time of the scram signal (14 seconds) given by low water level for this incident. 5.3.1 Description of Transient The sequence of events is as follows: 1. While operating at full power the feedwater pumps trip and the feedwater flow decays to zero in 2 seconds. 2. Reactor power and recirculation flow remain constant. In the transient without scram analysis" the power drops off after about 8 seconds due to the temperature rise in the reactor coolant, but in the present analysis the power was held constant to obtain conservative results. 3. Turbine throttle valve remains open so that steam flow is a function of reactor pressure. As the pressure changes little, steam flow remains nearly constant throughout the transient. 5.3.2 Results of Analysis Fig. 5.2 shows the voided, unvoided, and indicated water level as a function of time from the start of the transient (time zero) to 20 seconds. All water levels are plotted as a change from the normal level each occupies at time zero. The voided level is obtained directly from the computer printout as the mixture level of the steam dome and separator volume (Volume 3 of Fig. 5.1). The unvoided volume is cal-culated from the sum of the bubble and liquid masses of the same volume 1395 055 3_3
h I I l i I l l 0 _2 Voided Level A 4 ~ i Unvoided Level -6 y -8 g h-10 }- Indicated Level j -12 = m r_- = -== Scram Signal .c ( r. 6 -14 -16 -18 -20 -22 -24 I I I I I O 2 4 6 8 10 12 14 16 18 20 Time, seconds Fig. 5.2 - Voided, Unvoided, and Indicated Level, Loss of Feedwater Flow 1395 056 s-to
with the density taken as that of saturated liquid. The indicated level is the mixture levelin the standpipe (Volume 7 of Fig. 5.1). It can be seen that the voided, unvoided, and indicated levels all drop initially, as expected. As the flow of fresh feedwater has.been shut off and full power heat continues to be generated in the core, the temper-ature of the water in the reactor vessel and in the external loop (through mass and energy transport) increases and expands (see Fig. 5.3 for density variation), counteracting the effect of mass decrease in lowering the reactor level. As shown in Fig. 5.2, under the assumed conditions (such as no reactor scram) the voided level reaches a low of -4 inches and then turns up. The indicated level initially drops more precipitously due to sudden changes in pressure differentials between the reactor vessel and the standpipe and reaches an indicated level of -16 inches before turning up towards the true voided level. A low level reactor scram signal would be given when the indicated level drops to -12 inches at 6 seconds. 5.3.3 Conclusions 1. For the loss of feedwater incident the drop in true (unvoided) water levelis less than a simple reactor inventory mass balance would indicate, because of the decrease in water density following the loss of feedwater. 2. The indicated drop in water level during the initial period of the transient is greater than the actual drop, confirming the adequacy of the safety system. 3. Based on this investigation, a low water level scram signal, leading to corrective action, is given at an earlier time (at 6 seconds) than calculated in a previous study 28 (14 seconds), confirming the adequacy of the safety system set points as shown in the previous study.'8 5.4 LOSS OF COOLANT In a loss of coolant accident the high pressure emergency core spray starts whenever the indicated reactor water levelis -12 inches or less. The low pressure alternate core spray startt flowing into the reactor vessel when the water level falls below -12 inches, the contain-ment building pressure exceeds 5 psig, and the reactor pressure is below 150 psig. 1395 057 5-u
t I i i l l 1 Water in Intet Plenum 46 Water in Recirculation Loop / M&J l 44 ~ l W' * * ' I " ' * P * ' 9I" 43 and downcomer n Re ._g 42 E O 41 40 39 38 t I O 2 4 G 8 10 12 14 16 18 20 Time, seconds Fig. 5.3 -Variation in Water Density, Loss of Feedwater 1395 058 3_u
The ability of the levelindicator system to activate these sprays in time to protect the core from damage due to effects of loss of coolant type incidents has been investigated as part of a study of the adequacy of the LACBWR Emergency Core Cooling System." There, a RELAP model of the LACBWR, which included the standpipe and was similar to the RELAP model constructed for the present studies, was used to de-termine the times when a signal was given for the high or low spray flow and when the core would start to uncover. A table from that study, showing signal times and other parameters, is reproduced here as Table 5.5. The study of the Emergency Core Cooling System" established that the setpointe of the safety system were adequate, according to AEC criteria, to protect the reactor plant from damage as a result of a loss of coolant accident. As the model used in that investigation took into account water density changes and void swelling effects, no further analysis is required. 5.5 GENERATOR LOSS OF LOAD (PRESSURE INCREASE) The most severe increase in reactor vessel pressure occurs after a generator loss of load, such as the one described in Section 4.1 of Reference 16. The transient assumes an instantaneous total loss of steam flow to the turbine. In the previous analysis"it has been assumed that after a 15 psi rise in pressure the turbine bypass valve opens to relieve the rise in reactor pressure, but in the present analysis, in order to obtain larger pressure rises, it is assumed that the valve does not open and that the buildup of pressure continues. 5.5.1 Description of Transient The sequence of events is as follows: 1. While operating at full power there is a loss of generator load which in turn results in a loss of steam flow to the turbine, such that steam flow out of the reactor vessel decreases to zero within 0.01 second. 2. A scram signal causes the recirculation pumps to be cut back to 80% of full flow. This causes a reduction in reactor power to about 65% of design level within a few seconds, as shown by the curve of Fig. 5.4. In the present 1395 059 5-3 ~^ --
A
- 5. LACBWR LOSS OF COOLANT ACCIDENT s
In order to determine the clad temperatures in the LACBWR after a LOCA, the period after the break was divided into a number of phases, each requiring a different analytical model to represent the dominant modes of heat and mass transfer. Three different phases were modeled representing coolant blowdown, rod heatup, and fuel assembly heatup. The details are given in Sections 5.1, 5.2, and 5.3. Core spray flow rates are discussed in Section 5.4. Results of the analysis are given in Section 6. 5.1 BLOWDOWN ANALYSIS This analysis considers the processes which occur while the re-actor vessel empties after a coolant line has been severed. (The time of severance of the line is designated as time zero, in all the analyses which follow.) The blowdown analysis yields the following parameters which are utilized in subsequent analyses (Sections 5.2 and 5.3):
- 1. Reactor vessel pressure versus time
- 2. Coolant inventory versus time
- 3. Core flow rate versus time
- 4. Water level in reactor vessel versus time
- 5. Time elapsed before core uncovers
- 6. Containment pressure versus time The water level calculations yield the time for initiation of the H.P. core spray signal (-12 inches). Vessel pressure (<150 psig) and containment pressure (25 psig) determine the time the LPCS will enter the reactor. The time elapsed before the core uncovers is significant inasmuch as the cooling mechanisms in the core changes at that time b-(Section 5.3). The coolant inventory and flow through the core deter-
'4 1395 060
j mine the heat transfer mechanism applicable during the initial blowdown h ] period. Inasmuch as the blowdown of the LACBWR is extremely rapid for any but the smallest size recirculation line breaks, film boiling has been assumed to start at time zero, regardless of what critical heat fluxes were calculated by the blowdown code (RELAP 3). The clad temperatures existing in the high power fuel assembly were not taken from the computations of the blowdown code, but rather were calculated 1 by means of separate analyses, described in Sections 5.2 and 5.3. 5.1.1 Method of Analysis The blowdown sequence was simulated with the RELAP 3 com-puter program.7 For this code the reactor and associated systems are divided into discrete volumes. These are connected with junctions. The program calculates the flows between volumes and the thermo-dynamic conditions within each volume. Average fuel rod temperatures are calculated for the volumes representing the core. Pumps and valves are also included in the simulation. Fuel rod heat transfer coefficients are calculated by one of a set of seven equations. The equation used at any given time and location in the core is a function of the water quality in the surrounding volume. Critical heat flux is determined from one of six correlations, depend-ing on pressure and mass flow. As mentioned above, these critical heat flux calculations were not utilized in the determination of the high power fuel element cladding temperatures, inasmuch as immediate film boiling was assumed. One modification was made to the RELAP 3 Code; the nucleate boiling heat transfer coefficient was limited to a maximum value of 10,000 Btu /hr-ftz-F, irrespective of what the code calculated. This eliminated a convergence problem that arose as the flow in the core reversed after a recirculation line break. The analysis was performed in accordance with the interim criteria requirements stated in Appendix A, with the exception of Items 1 and 2, which are directed specifically toward GE plants. These items do not apply, inasmuch as the LACBWR plant does not utilize jet pumps, and the blowdown does not result in lower plenum ilashing. / 5-2 = I i i
s 1 14 civI r-----------------------l # I i l' s vehe l ^ Stee. pie. T I so e 7' l I I 21' l l o 5-l F l l 0 5 1 I Steam separator l 10 g 8 l <i 4' l l Down i l l 3 l = i l 2 15 l l 5 I + I l 1 l 19' g l i e ,,1 I I 1 l l l 4 l Inlet plenum 10' l e 10 9 e l l l 12' 1A e 1A -+ l Recirculation Pump Recirculation I l g. o 18' dscharge auction l pg, I l l 17' 16* 13* ~ l 18 IB + l Recircn4stion Pump RecMh l
- ch*'s' 15-l l
j L______________________J j Volume l o 1' Junction i i 3 h .R Fig. 5.1-Blowdown Analysis Model, Recirculation Discharge Line Double-Ended Break 5-4 1395 062
Analyses were made for the following accidents.
- 1. Double-ended break of the 20-in. diameter recirculation pump discharge line.
- 2. Double-ended break of the recirculation pump suction line.
- 3. Single-ended breaks in the recirculation discharge line 2
with areas of 1. 0, 0. 5, 0. 25, 0.1, and 0. 05 f t.
- 4. Steam pipe break.
All breaks were assumed to occur when the reactor is operating at the full power of 165 Mw(t). A simultaneous loss of off-site power was also assumed. 5.1.2 Description of Models Double-Ended Recirculation Pump Discharge Sixteen volumes and 21 junctions are included as shown in Fig. 5.1. These are described in Tables 5.1 and 5.2. Details of the calcu-lations are given in Appendix B. Volumes 1 through 3 represent the core, with power fractions of 0.474, 0.348, and 0.178, respectively. These fractions were determined from a typical axial power distribution, as shown in Appendix B. Only the average channel was modeled for the blowdown analysis. All power is generated in the fuel pellets. To represent the rod heat transfer, the fuel is divided into four nodes, the clad into four nodes, and the gap is a single node. Properties are presented in Table 5.3. Initial flows and junction qualities were set for steady-state operation. The break was assumed to occur at the junction of the 1B 2 recirculation discharge pipe of 1.795 ft cross sectional area and the 2 inlet header of 1.141 ft cross sectional area (see Fig. 5.1 for location of break). The full outflow was diverted to the containment vessel beginning at time zero. Power was maintainrxi at 165 Mw(t) for 0.5 seconds and then re-duced to the decay heat value linearly over the next 0.5 seconds. This simulated the effect of voids on the reactivity. Decay heat continued to be generated as calculated by the Shure correlation, (ANS Standard) allowing 5-3 1395 065
h TABLE 5.1 - VOLUME GEOMETRY AND INITIAL CONDITIONS
- Volume, Height, Lowest
- Pressure, Temperature, Number Region it ft Elevation, it psia
- F 8
1 Lower core 33.5 2.5 656.75 1304.56 577.6 2 Middle core 25.7 1.92 659.25 1303.28 577.5 3 Upper core 33.5 2.5 661.17 1302.13 577.3 4 Lower plenum 291.9 9.0 647.75 1307.33 563.2 5-Above core 280.2 6.9 663.67 1300.63 577.2 6 Steam dome 680.4 14.3 670.56 1299.49 578.0 7 Steam line 36.7 43.0 634.0 1298.00 577.8 8 Downcomer 324.0 13.5 652.92 1301.47 577.3 9 1A recirculation suction 113.9 28.0 626.0 1303.98 563.2 10 1A recirculation discharge 87.9 23.0 626.0 1316.03 563.2 11 Turbine 100.0 10.0 634.0 1296.00 577.7 [ 12 IB recirculation c suction 104.1 28.0 626.0 1303.98 563.2 W 13 IB recirculation c-, discharge 98.7 23.0 626.0 1316.03 563.2 { 14 Containment 264,160.0 139.0 617.5 14.7 80.0 15 Standpipe, lower 0.1702 8.3 657.0 1301.02 577.3 16 Standpipe, upper 0.243 11.9 665.2 1299.62 578.0 ? w
TABLE 5.2 -JUNCTION GEOMETRY AND INITIAL CONDITIONS Volumes Elevation,
- Inertia, Number Connected it Flow, Ib/see it-'
1 4-1 656.75 2986.0 0.267 2 1-2 659.25 2986.0 0.165 3 2-3 661.17 2986.0 0.165 4 3-5 663.67 2986.0 0.242 5 5-6 670.56 169.7 0.175 6 6-7 677.00 169.7 71.7 7 7-11 634.00 169.7 108.7 h 8 5-8 663.67 2816.0 0.43 9 8-9 654.0 1408.0 13.0 10 0-9 654.0 84.85 0.0 11 9-10 626.0 1493.0 ' 31.4 12 10-4 649.0 1493.0 8.9 13 8-12 654.0 1408.0 11.5 14 11-0 635.0 169.7 0.0 15 0-12 654.0 84.85 0.0 16 12-13 626.0 1493.0 31.5 17 13-14 649.0 1493.0 12.0 18 4-14 649.0 0.0 1.83 19 8-15 657.0 0.0 1590. 20 15-16 665.3 0.0 491. 21 16-6 677.15 0.0 1514. 1395 065 g 5-6 L
TABLE 5.3 -MATERIAL PROPERTIES, RELAP-3 CODE Specific
- Density, Temperature, Conductivity,
- Heat, Ib/ft3 F
Btu /hr-f t-F Btu /lb 'F Fuel 649.9 500 3.780 0.0690 1100 2.473 0.0745 1700 1.818 0.0775 2300 1.494 0.0795 3500 1.368 0.0990 4300 1.800 0.1150 5100 2.736 0.1300 Helium 0.0050 800 0.092 0.124 1520 0.220 0.124 3140 0.373 0.124 Clad 500 400 11.0 0.12 800 13.0 0.13 1200 15.0 0.14 1600 18.0 0.16 3000 18.0 0.16 139.5 066 4 -4 5-7
also for the decay of activation products and with an added uncertainty g. factor of 20%. The power curve is shown in Fig. 5.2 The pump head was reduced to zero immediately as the flow through the pumps quickly exceeded their tgacity. The steam flow to the turbine was shut off instantaneously when the pressure dropped to 1000 psia in the steam dome. This is a con-servative assumption, inasmuch as the turbine throttle vaive actually closes sooner. Feedwater flow was assumed to drop to zero at the time of the break. The emergency core spray and the shutdown condenser (brought into service at 1000 psia) were not included in the model. This repre-sents an added conservatism to the analysis. Double-Ended, Recirculation Pump Suction Line The same model was used except that the break was at the junction of the IB suction header and the reactor outlet plenum. Single-Ended, Recirculation Pump Discharge Line Two models were used for th9 single-ended breaks. The first was the same complete model as for the double-ended recirculation supply line break, except that junction 17 connected volumes 13 and 4, and junction 18 was the only break. Break areas of 1. 0, 0. 5, 0. 25, and 2 0.1 ft were investigated with this model. In addition, a simpler model was developed for studying small breaks that have a slower effect on the system. The simpler model reduces the computer time required to carry the investigation out for longer time periods. The model, shown in Fig. 5.3, reduces the 16 volumes of the larger model to 8. The standpipe was included for those runs where the time of the high pressure spray trip signal was required, but it was deleted for long runs. Break sizes of 0.25 and 2 0.05 ft were investigated with the simplified model Break areas smaller than 0.05 ft were not analyzed with the RELAP Code as even 2 with the simplified model the computer time would have been prohibi- ~ tive. Required data for these breaka has been determined from extrap-j olations of the larger break results. G }.595 067 5-8 l
O 20 l to ^ ( j s i Vy 3 -10 L g 20 0 0.2 0.4 0.6 0.8 1.0 Time, see Fig. 5.10 - Differential Pressure across Stand Pipe following Recirculation Line Double-Ended Break on Discharge Side 1395 068 ~ g 5-18 A0012
conditions (normal operating water level) the pressure differential be-tween the centers of the two volumes is about 1.4 psi. The pressure differential history after a discharge line break, as calculated by the RELAP 3 Code is shown in Fig. 5.10. The erratic pattern of the curve is due to its sensitivity to small changes " flow rates. A drop in level of 12 incc us, sufficient to give a H.P. core spray signal, is equivalent to a reduction in differential pressure of 0.3 p.si. As seen from Fig. 5.10 this occurs almost instantaneously. To be conservative, the initiation of the H.P. core spray signal for this break has been taken to occur at 0.14 seconds when the AP curve becomes negative for the first time. The low p ressure spray activates when the reactor water level drops to -12 inches and the containment pressure reaches 5 psig. Low pressure spray enters the reactor vessel when the vessel pressure de-creases to 150 psig. For the discharge line break, this occurs at 5.5 seconds, and for the suction side, at 6.5 seconds. In both cases, as shown in Fig. 5.5, the reactor pressure decreases rapidly to the con-tainment building pressure, 47 psig. The containment 'ouilding pressures were calculated by means of the CONTEMPT Codc8 and are given in Tarle 5.4. Single-Ended Breaks Results of the single-ended breaks are shown in Figs. 5.11 through 5.16. Reactor pressure decays more slowly, in proportion to the break size (Fig. 5.11). None of the breaks voids the core faster than the double-ended discharge break. The core inventory for the 0.5 ft 2 reak (Fig. 5.12) shows a rise from 5 to 9 seconds, then a rapid drop. This is a result of major flow changes. The core is a small part of the whole system and is strongly affected by changes in other parts of the system. The 0.25 break shows the start of a similar effect at about 10 seconds. Heat transfer coefficients (Fig. 5.13) are higher than for the double-ended break, especially after six seconds, and they remain higher because of the larger water inventory in the system. The water level above the active core is shown in Fig. 5.14. The 2 analyses for the 0.5 and 0.25 ft breaks used an improved steam sepa-ration model that predicted higher levels. The results for the 0.1 2 and 1.0 ft breaks do not reflect this improvement and are thus more conservative. A similar effect is seen on the core inventory plot (Fig. 5.12). The two smaller breaks were not carried out to the point where 1395 69 L
______=.__m-- i TADLE 5.4 -
SUMMARY
OF BREAK CHARACTERISTICS Decay Heat Time of Time Top Time of Decay Heat when L.P. Maximum H.P. Spray of Core L. P. Spray When Core Spray Containment Break Size
- Signal, Uncovers,
- Flow, Uncovers, Starts, sec Pressure it*
sec see sec psig 2.93 0.14
- 1. 6 6
7.3 6.8 47 1.0 0.3
- 2. 6 13
- 7. 0 6.2 41
- 0. 5 0.04
- 8. 9 26
- 6. 6 5.4 37 0.25 0.18 15 41
- 5. 9
- 4. 9 33 0.10
- 0. 2 39*
13 0*
- 5. 0 4.1 27 0.05
- 2. 2 78*
260*
- 4. 4 3.6 23 0.025 N.C.t 156*
520'
- 4. 0
- 3. 0 19 0.01 N.C.
432* 1,430* 3.1
- 2. 0 13 O.005 N. C.
960* 3,200*
- 2. 5
- 1. 6 8.3 0.00125 N.C.
13,000* 45,000* 1.15
- 0. 8 0
0.031 <= = 0 0 ro u, C
- Estimated values y
a f Not calculated T E 9
- . s 3.2 INSTRUMENT ERROR The water level within the reactor vessel is measured by on insulated water columr, which is connected to nozzles in the reactor vessel wall; these nozzles are iocated at elevations of 657 ft - 0 in, and 677 ft - i<75 in. The lower vcssel nozzle is in communicatien with the annular downcomer region at opproximatelv the some elevation as the Ectrom of the core. The top vessel nozzle is in communicatio, with the vessel steam dome, downstream of the steam dryers in the vessel head. The water level in the insulated water column outside of the reactor gives on indication of the equivalent height of soturated water within the reactor vessel above the lower nozzle.
The error in the indicated woler level is made up of errors due to (a) the difference be-tween the tempc roture of water in the column onJ the acter within the ieactor.essel, (b) the pressure drep across the steam dryers, ic; the pressure drop in 'ne downcor"er o'c' (d) the occuracy of the instrumentation used to rreasure and transmit the differential pressure in the insulated-water column. The indicated value corresponds to the un-soided water level within the vessel so that errors due to void content do not have to be accounted for. The temperature difference between the insulated-water column and the reactor vessel water is estimated to be 10 F or greater. This density difference con result in a meos-urement error of 1.1 in or greater. This is a negative error, i.e., one which gives on indicated value of water level lower than the octual value. The pressure drop across ths team dryers during full power operation is a maximum of 4 in. of water (see pg. 4-49 of the Sofeguards Report). This pressure drop would be re-flected as a positive error in water level measurement, i.e., the indicated level would be higher than the octual level by a maximum of 4 in. This error will vary with the steam flow rate through the steam dryers, or with reactor power level, from zero to o maximum of 4 in. O D ""E
- E N W n 1395 07!
a e M J JL XhL ll-3-26
9 se The pressure drop in the dowacomer et full power is 0.2 ft, or 2.4 in. (see Sec. 4.6.9.1 (4) of the Sofeguards Report). This pressure drop will give o negative error in the indi-cated water level, i.e., the indicated water level con be lower than the octual by 2.4 in. This error is primarily a function of the recirculation flow rate. Over the operating range, it should vary from o minimum value corresponding to 40 percent of full flow to 2.4 in. The occuracy in the measured value of water level within the insulated-water column, including the instrumentation and equipment used to calibrate this instrumentation, is equal to i 0.40 percent of scale, or t 0.24 in. The combined occuracy for indicated salues of unsoided water level within the reactor vessel is summarized below : error due to temperature difference of insulated- - 1.1 in. water column steam dryer pressure drop +4 in. -2.4 in. downcomer pressure drop differential pressu e inst,umentation
- 0.24 in.
r +0.510.24 total The res ;1t is equal to on error of + 0.74 in. to - 0.26 in. Thus, the indicated value of water level could be higher than the oc'uol un <oided water level within the vessel by 1/4 in. to 3/4 in. With this error, the minimum raorgin betwee, the octuct water level at the tc u-level soom :>oin* cra the minim;m o Howable level cor respondirg to on un-voided water level of the top of 'he foel element shouds would be 8-l/2 ir, 3.3 INSTRUMENT RESPON5E TIMES The response time of the water level measurerrect instrumentation ts essentially equal to the respc-se time of the differcntial pressure transmitter, or 0.2 sec. Thur, within 0.2 sec of ter existence of a low levei condition, a low-level signal would be transmitted to the safety system. This signal will initiate o reactor scram and will initiate closure of the main steam line isolation volve. A delay time of 0.5 sec is essumed for both of these actions, although the octual delay time is expected to be less. hb].h Ql} l 1 e ll 27}}