ML20050A890
| ML20050A890 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 02/24/1982 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-05.A, TASK-RR LAC-8113, NUDOCS 8204020369 | |
| Download: ML20050A890 (23) | |
Text
g,/ t v,
DlDA/RYLAND k
[k COOPERAT/VE PO BOX 817 2615 EAST AV SOUTH
- LA CROSSE. WISCONSIN 54601 (608) 788-4000 Feb rua ry 24, 1982 In reply, please refer to LAC-8113 DOCKET NO. 50-409 U. S. Nuclear Regulatory Commission ATTN:
Mr. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation d
ID Division of Operating Reactors D
p
Washington, D. C.
20555
\\
RECGlW:,'y
'l \\
SUBJECT:
DAIRYLAND POWER COOPERATIVE d
LA CROSSE BOILING WATER REACTOR (LACBWR) 2 Wa 02 ; 82m-O SEP TOPIC III.S.A
{j a ama,7
/
g EFFECTS OF PIPE BREAK ON STRUCTURES,
, 8Eara SYSTEMS AND COMPONENTS INSIDE CONTAINMENT
'e og.
REFERENCE:
(1) DPC Letter, LAC-7751, Linder to Eisenhut, D
dated August 19, 1981.
(2) DPC Letter, LAC-8067, Linder to Crutchfield, dated February 4,1982.
Gentlemen:
Enclosed find the Safety Evaluation Report (SER) for Effects of Pipe Break on Structures, Systems and Components Inside Containment (SEP III.S.A) which we have prepared for the La Crosse Boiling Water Reactor.
Our letter, Reference 1, identified topics for DPC to submit for NRC evaluation.
The subject topics were listed in the schedule submitted with Reference 1.
This schedule was revised in Reference 2.
If there are any questions regarding this letter, please contact us.
Very truly yours, DAIRYLAND POWER COOPERATIVE et ; d/
E re d d L L Frank Linder, General Manager FL:JDP:eme cc:
J. G. Keppler, Regional uirector, NRC-DR0 III NRC Resident Inspector b
o WP-1 l
g 8204020369 820224 h
PDR ADDCK 05000409 P
O O
w I
LA CROSSE BOILING WATER REACTOR SYSTEMATIC EVALUATION PROGRAM SAFETY EVAULATION REPORT TOPIC III.5.A EFFECTS OF PIPE BREAK ON STRUCTURES, SYSTEMS AND COMPONENTS INSIDE CONTAINMENT I.
INTRODUCTION The safety objective of the Systenat'ic Evaluation Program (SEP) Topic III.S. A, " Ef fects of Pipe Break on Structures, Systems and Components Inside Contai nment," is to assure that pipe breaks would not cause loss of l
systens required to safely shutdown the reactor.
The review includes consideration of redundant systems capable of performing accident mitigation and plant shutdown should a specific systen be damaged by a high energy pipe fracture in its vicinity.
II. REVIEW CRITERIA The review was based upon an outline provided by the Nuclear Regulatory Commission (Reference 1).
This outline provided three acceptable approaches to evaluate the effects of postulated breaks in fluid system piping inside containment upon components of essential systems.
Three methods were detennined to be:
a fully mechanistic approach which o
utilizes stress analysis for postulating break locations, i
an effect oriented approach which postulates breaks in the immediate vicinity (i.e., most critical locations) of the safety-related equi pment.
a simplified mechanistic approach which postulates breaks at terminal
{
ends, at each pipe fitting and at each weld.
III. REVIEW METHOD The ef fects oriented approach was selected for this review at LACBWR.
l Each systen inside containment was reviewed in accordance with the criteria listed in Reference 1 and supplemental guidance given by the NRC piping run was evaluated if:
it contained pipes greater than 1" in diameter 1
o 1
it was in service or pressurized at least 1% of the time.
o it had a maximum operating temperature over 200 F and/or had a maximum o
operating pressure over 275 psig.
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l
4 s.
i Once a piping run was selected for evaluation fractures were postulated at terminal ends and points of closest approach to components of essential equipment to determine the effects of pipe whip or impingement.
4 Damage was assumed where equipment was in the path of impact or smaller diameter piping was in the path of impact. The equipment was assumed to be rendered inoperable and redundant systems or mitigations when necessary were discussed.
The US NRC reviewed with DPC and DPC's consultant the draf t evaluation 1
during a site visit in August of 1979 and suggestions were incorporated.
)
IV.
SYSTEM SELECTION All applicable systens within the containment were reviewed.
Table 1 lists the thirty piping systems identified. Twelve systems met the
]
criteria for high energy systems for at least part of their piping.
These systems were:
a forced circulation o
i main steam o
=
i
)
alternate core spray o
o
)
seal injection system o
i control rod drive hydraulic systen o
control rod drive effluent systen o
i shutdown condenser o
i o
boron injection decay heat o
i purification o
i I
l I
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-2 l
i
~
..._-_._, _._ _ -, _ _,, _ _._ o
O The essential systems having components inside contain1ent include:
High Pressure Core Spray - both pumps and associated piping o
Alternate Core Spray - piping and check valves o
Boron injection System - boric acid storage tank and associated o
piping Manual Depressurization System - valves, piping and the shutdown o
condenser Control Rod Drive Hydraulic Scram Accumulators o
Essential instrumentation - reactor vessel water level reactor o
vessel pressure Breaks inside containment may result in a loss of coolant accident or a lesser situation requiring only nonnal plant shutdown.
CONCLUSIONS Each point of posutlated break was addressed in Table 2.
The criteria of Reference 1 was utilized in looking for a mitigating action or redundant system. A recent draft of a safety evaluation on this topic for another f acility added new consideration:
loss of power and a single failure proof redundant system or nitigating action. These are added as applicable.
The piping system runs not identified below (but included in Table 2) are satisfactory without further consideration.
1.
Boron Injection System 1.1 A break at or near the biological shield could damage the containment ventilation exhaust damper operators.
1.2 A break near the purification platform could damage the reactor water level system reference.
The potential breaks would not prevent cold shutdown of the reactor from being achieved.
The enclosed sketch indicates the boron injection high pressure core systems. These lines are being analyzed for seismic events and DPC is near completion of the seismic analysis and design and installation of required actions and supports.
This piping analysis and restraint gives a higher degree of confidence to the lines in this system.
The inservice inspection program for Class I Systems (ASME Section XI) includes 100% of the boron inject system.
Also, the welds in the system have been examined according to the augmented ISI Program approved for LACBWR which give a high degree of dSsurance to these piping runs retaining their integrity.
We conclude that the boron system line is acceptable and no further consideration is necessary.
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-3
2.
Seal Injection Systen, Main Steam System, Control Rod Effluent System 2.1 A break in the seal injection system piping piping near its terminus at the seal injection pumps or a point in the piping could possible cause damage to control rod drive mechanisms.
l 2.2 A break in the main steamline at the point of existing the lower reactor cavity could possibly damage control rod drive mechani sms.
2.3 A break in the control rod effluent piping near the control rod 4
drive mechanisms could cause damage to the mechanisms.
While the high energy line breaks defined above have the potential to i
cause sone damage to the control rod drive mechanisms, it is believed that the damage, if any, would be a partial failure to insert control rods.
In any event, the boron injection system would still remain available for reactor shutdown. While the boron injection system has not been designed to be single failure proof, it has been designed i
with sufficient capacity to place the plant in a shutdown condition i
with all control rods removed. The potential damage to the decay heat line from pipe breaks in the decay heat piping or main steam piping could affect the containment isolation valve.
OPC will install a manual valve in the turbine building reheater area which will be closed by procedure following an incident of this nature.
l Based on the above evaluation, we conclude that the analysis for high energy line breaks with the potential to damage the control rod drive I
mechanisms is acceptable for the La Crosse Boiling Water Reactor.
i 1-REFERENCE 1: Letter Davis, NRC to McEwen, K M C, Incorporated, i
Dated July 20, 1978.
1 WP_1
_4 t
i
Page 1 of 3 TABLE 1 LA CROSSE BOILING WATER REACTOR (LACBWR)
PIPE BREAK INSIDE CONTAINMENT (SEP TOPIC III 5.A)
SUMMARY
OF HIGH ENERGY LINE CLASSIFICATION
,R i
j If. SERVICE OR MAXIMUM MAXIMUM l-SYSTEM CONTAINS PRESSURIZED OPERATING l
OPERATING HIGH l CON TAINMENT INSIDE PIPES AT LEAST TEMPERATURE PRESSURE ENERGY
> 1" D I A.
1% OF TIME
> 200 F.
l
> 275 PSIG SYSTEM q
FORCED CIRCULATION X
X X
l X
YES t
I IFEE0 WATER X
X X
X YES l
liiAIM STEAM X
X X
l X
YES l
l l
l l AL TERNAi E l
l CORE SPRM X
X X
l X
YES -
UP TO l
HIGH PRESSURE l
CORE SPRA(
X X
X X
YES I
HIGH PRES 5t1RE l
l SERVICE L'ATER X
X NO l
l SEAL INJECTION l l
SYSTEM X
X X
YES I
l l CONTROL AIR X
X l
NO l
STATION AIR l
X X
N0 I
DEMINERALIZED l
WATER X
X N0 l
l CONTROL ROD l
DRIVE HYDRAULICS X
X l
X YES
e Page 2 of 3 TABLE 1 LA CROSSE B0ILING WATER REACTOR (LACBWR)
PIPE BREAK INSIDE CONTAlfiMENT (SEP TOPIC 111 S.A)
SUMMARY
OF HIGH ENERGY LINE CLASS!FICATION I
I IN SERVICE OR MAXIMUM MAXIMUM SYSTEM CONTAINS PRESSURIZED OPERATING l
OPERATING HIGH INSIDE PIPES AT LEAST TEMPERATURE PRESSURE ENERGY C0f4TAlhMENT
> 1" DIA.
1% OF TIME
> 200* F.
l
> 275 PSIG SYSTEM l
l CONTROL ROD DRIVE l
NITROGEN X
X NO l
l l CONTROL ROD l
l EFFLUENT ISYSitM X
X X
l X
YES I
SHUTUOWN YES CONDENSER l
UP TO CONDENSATE STEAM &
I AliD l
CONDENSATE STEAM LINES l
X X
X ISOLATION l
l VALVES.
l l
. FUEL STORAGE WELL COOLING l
SYSTEM X
X N0 l
RETENTION TANK l
1RANSFER LINES X
X N0 l
RESIN SLUICE l
LINES X
N0 l
BUILDING DRAIN l
LINES X
X NO l
CONTAINMENT l
VENTILATION SYSTEM X
X l
N0 l
l CONTAINMENT VENT l
HEADER X
X N0 l
l l
Page 3 of 3 TABLE 1 LA CROSSE BOILING WATER REACTOR (LACBWR)
PIPE BREAK INSIDE CONTAINMENT (SEP TOPIC III 5.A)
SUMMARY
OF lilGH ENERGY LINE CLASS!FICATION l
l l
IN SERVICE OR MAXIMUM MAXIMUM SYSTEM l
CON TAINS PRE SSURIZED OPERATING l
OPERATING HIGH INSIDE l
PIPES AT LEAST TEMPERATURE PRESSURE ENERGY l CON TAlllMENT l
> 1" DIA.
1% OF TIME
> 200 F.
l
> 275 PSIG SYSTEM l -- -
l l
l l
l lHVDRAULIC VALVE l ACCUMULATOR l
- SYSTEM X
X N0 l
lOlfEkHEA0 l
l STORAGE TANK SYSTEM X
X l
N0 I
COMPONENT
,COULING l
WATER SYSTEM X
X N0 l
HEATING STEAM, l
NO PART OF
. VENTILATION l
SYSTEM X
X l
l SillELD COOLING l
SYSTEM X
X N0 l
IBORON l
- YES, INJECTION X
X X
UP TO l
CHECK VALVES l
l DECAY HEAT X
X X
l X
YES 1
l PURIFICATION X
X X
X YES I
FIRE HEADER X
X l
NO l
gg[g{gENRAY,l X
l N0 1
l l
Page 1 TABLE 2 l
l l
SAFETY-RELATED EQUIPMENT EFFECT OF BREAK ON REOUNDANT SYSTEMS OR l SYSTEM j PIPING RUN I
POSTULATED BREAK IN VICINITY SAFETY-RELATED EQUIPMENT MITIGATING ACTIONS Boron Parallel Terminal break at No safety-related equip-None None required.
Inject-I sol ation forced circulation ment of small pipe size Principal shutdown ion Valves 60 system discharge (1 1/2-2 1/2' ) is located mechanism (control l
002 & 60 header or at in the lower reactor rod scram function) l 006 to dis-biological shield, ca vi ty.
No transmitters is unaffected. The l
charge header or cable trays are fractured line would l
of forced located there.
represent a 2 1/2' l
isolation LOCA.
l sy sten.
l I
l l Boron Parallel Terminal break Containment building Potential mechanical or No redundant isola-l Inject-l Isolation outside of the ventilation discharge jet impingement damage tion system would be lion Valves 60 biological shield.
isolation dampers.
to operators of the available. A piping l
002 3 60 Containment Building restraint on the j
006 to dis-ventilation discharge boron injection line l
charge header dampers to reduce the length of forced free for movement isolation would eliminate the system.
risk of damage to the ventilation damper operators. The fractured line would l
represent a 2-I/2' l
LOCA.
Boron Parallel Break while passing.
Reactor water level Possible damage to small No redundant water Inject-I solation puri fi cation reference leg including piping off of reference level instrumentation ion Valves 60 platform.
standpipe heaters, leg.
would be available.
1 002 3 60 A ripe rastraint will l
l 006 to dis-eliminate the poten-charge header tial risk. The frac-of forced tured line would isolation represent a 2 1/2' syst em.
LOCA.
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Page 2 TABL_E 2 a
l l
SAFETY-RELATED EQUlPMENT EFFECT OF BREAK ON REDUNDANT SYSTEMS OR lSVSTEM l PIPING RUN POSTULATED BREAK IN VICINITY SAFETY-RELATED EQUIPMENT MITIGATING ACTIONS l
l l Boron Parallel Terminal break at Isolation valves for the Loss of the high pressure The redundant system l Inject-1 Isolation either isolation boron injection system.
core spray system.
(Alternate Core Spray l ion Valves 60 valve or both.
Manual Depressuriza-l 002 3 60 tion) remains j
006 to dis-available. The l
charge headers fractured line would l
of forced represent a 2 1/2' l
circulation l
LOCA.
l sy stem.
l l
l l
I I
- High Pump dischargei Terminal break by Boron injection syst e loss of boron injection The reactor would be Pressu re check valves
) pumps.
piping and one of three system and loss of one shutdown by the re-1 Core l to reactor reactor water level reactor water level dundant shutdown Spray
~ vessel.
I condensate pots.
channel.
system:
The control rod hydraulic scram.
-j The otte? two reactor l
j water level channels j
l would provide indica-l l
l l
tion. The fractured l
j l
pipe would represent l
l l
a 1 1/2' LOCA.
I i
I HP-1
Page 3 TABLE 2 l
l l
a SAFETY-REL ATEU EQUIPMENT l
EFFECI 0F BREAK ON REDUNDANT SYSTEMS OR l SYSTEM l PIPING RUN l
POSTULATED BREAK l IN VICINITY SAFETY-RELATED EQUIPMENT MITIGATING ACTIONS i
1 l
jHigh l Pump discharge Break at mezzanine Low pressure safety injec-Loss of low pressure The low pressure core' l Pressure j check valves level by the boron tion 30 psi differential safety injection differ-spray would probably l Core l to reactor injection tank.
pressure transmitter.
ential pressure trans-receive an operate l Spray i vessel.
Reactor safety channel mitter and loss of both signal on loss of
~
l l
transmitters reactor reactor safety channels transnitters, however l
l l
pressure 1 & 2 and air for reactor pressure.
the redundant system l
l activity monitors.
Alternate Core Spray /
l Manual Depressuriza-l tion is available and l
not affected by the pipe break. The reactor pressure l
l safety signals would j
not be required to i
l function due to the depressurization f rom i
l l
the LOCA.
[
l 1;High Pump discharge l Closest point of Equipment controlled or Loss of control or power The redundant shutdown l Pressure check valves
, approach to cable powered by cables in cable to: boron injection system (control rod l
Core to reactor tray #126.
tray #126.
pipe heating, boron tank hydraulic system) and Spray vessel.
Ieyel indication and the redundant l
l cunt.rol of the inlet emergency core cooling l
valves in the boron system (Alternate Core j
injection system. Loss Spray / Manual Depress-l of power supply to the urization Systen) l high pressure core remain available.
j spray pumps.
I i
l l
l I
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Page 4 TABLE 2 l
l l
, SAFETY-RELATED EQUIPMENT EFFECT OF BREAK ON REDUNDANT SYSTEMS OR l SYSTEM l PIPING RUN l
POSTULATED BREAK IN VICINITY I
SAFETY-REL ATED EQUIPMENT '
MITIGATING ACTIONS I
Hign l Pump discharge Tenainus at check Reactor pressure Loss of one indication of Primary system Pressure l check valves valve on pump transmitter 64-35-307.
reactor pressure in the pressure is available Core l to reactor discharge.
Control Room.
Loss of on the primary system.
[ Spray l vessel.
I signal to the narrow safety channels. The j
j range controller for the reactor pressure l
l Main Steam Bypass Valve, control provided by l
l the main steam bypass n
l I
I I
l Shutdown l Junction with Tenninus by Alternate core spray Damage to alternate core The main steam bypass lCondenserj main steam biological shield.
piping, main steam bypass spray piping, loss the valve would not be jRelief
, line to valve control transmitter transmitter.
required to function Valve parallel 53-25-001.
to mitigate a LOCA. A
would quickly result l
in a loss of primary j
system pressure. The j
redundant ECC System j
(High Pressure Core l
Spray supplied by the overhead storage tank) would remain avail-l able.
The ability of I
the Alternate Core l
Spray System to flood j
the Containment Build-ing would not be affected by this l
break.
I i
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Yage 5 TABLE 2 j
SAFE TY-RELATED EQUIPMENT l
EFFECT OF BREAK ON REDUNDANT SYSTEMS OR SYSTEM l
PIPING RUN POSTULATED BREAK IN VICINITY SAFETY-RELATED EQUIPMENT MITIGATING ACTIONS l
l l Shutdown l Junction with Point of closest Alternate Core Spray Damage to the check The redundant ECC l Condenser l Main Steam approach to the piping and check valves.
valves could cause a loss l system (High Pressure l Relief line to alternate core of primary coolant Core Spray supplied by.
lValue parallel spray check valves.
through both the broken the Overhead Storage l Header l isolation steam line and the Tank) would remain l
l valves, alternate core spray line available the ability l
l simultaneously. Analysis of the Alternate Core l
l l
for steam phase LOCA Spray System to flood j
l l
events demonstrated that the containment will j
l resulting cladding not be affected by l
l temperatures are less this break. Potential l
l than operating levels, loss of the Contain-l l
The total break cross ment isolation func-l section for this event is tion of the check j
less than for a double valves is backed up l
ended recirculation pipe by the motor operated l
fracture.
valves 38-30-001 and l
l l
38-30-002.
I l
l l
l Shutdown l Junction with One-half the length High pressure core spray loss of the high pressure The redundant ECC l Condenser l Main Steam fran the mezzanine system suction froa the core spray system.
system, Alternate jRelief l line to to the 701' level.
overhead storage tank.
Core Spray / Manual Valve parallel Depressurization would Header isolation remain available l valves.
throughout the l
incident.
l l
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Page 6 TABLE 2
.I l
l i SAFETY-RELATED EQUIPMENT l
EFFECT OF BREAK ON
' REDUNDANT SYSTEMS OR I
SAFETY-RELATED EQUIPMEf4T MITIGATING ACTIONS SVSTEM l PIPIf4G RUN POSTULATED BREAK IN VICINITY i
I l Shutdown L Junction with At the shutdown Shutdown condenser to The structures and piping None required.
Condenser Main Steam condenser isolation manual depressurization are of greater size and Relief line to valves 62-25-011
- valves, mass than the steam l l Valve pa ral lel and 62-25-001.
inlet lines. The manual l -: Header isolation depressurization valves 3 valves.
however, could be physically damaged. As
.l the break ';n the piping l
run takes the shutdown condenser out of service and accomplishes the depressurization it has no further af fect on l
that system.
I l
l Shutdown Condensate Terminus in at Shutdown condenser None, the shutdown None required, other jcondenser return line feedwater line.
condensate flow condenser's function is than shutting the l
parallel check tra nsmitter.
lost if this line breaks.
plant down.
l valves l
62-26-001 &
l 62-26-006 l
to the feed-l l water line.
1 WP-1
Page 7 e
TABLE 2 l
l l
l SAFETY-RELATED EQUIPMENT
]
EFFECT OF BREAK ON REDUNDANT SYSTEMS OR J STEM l
PIPING RUN POSTULATED BREAK '
IN VICINITY SAFETY-RELATED EQUIPMENT
_ MITIGATING ACTIONS i eal l Separate run Terminus into Control drive mechanisms Possible damage to the If the scram function S
l Injection to control i common header, wi re tray - #109,110,111.
control rod drive on some control rods l Sy stem rod drive mechanisms. Loss of the were lost, the boron l
l mechanisms.
wire tray which supplies injection system is l
l power to the control rad capable of reactor drives indicator, etc.
shutdown with all rods nonfunctional.
l
\\
l Seal Separate run Any location on Control drive mechanisms Possible damage to t,e If the scram function Injection.
to control header run up to wi re trays #109,110,111.
control rod drive on some control rods Sy stem rod drive and including mechanisms. Loss of the were lost, the boron mechanisms.
final branching.
wire tray which supplies injection system is l
power to the control rod capable of reactor l
drives indicator, etc.
shutdown with all rods.
j nonfunctional.
l l
l 1
l l
l Seal l Seal injection Terminus at seal Control drive mechanisms Possible damage to the If the scram function Injection. pumps to injection pumps.
wi re trays #109,110,111.
control rod drive on some control rods
- System forced circu-mechanisms. Loss of the were lost, the boron lation pumps wire tray which supplies injection system is j
except supply power to the control rod capable of reactor j
to control dirves indicator, etc.
shutdown with all rods l
l rod drives.
nonfunctional.
l l
1 1
I WP-1
Page 8 TABLE 2 l
l l SAFETV-RELATED EQUIPMENT EFFECT OF BREAK ON REDUNDANT SYSTEMS OR SYSTEM l
PIPING RUN l
POSTULATED BREAK IN VICINITY l
SAFETY-RELATEC EQUIPMENT MITIGATING ACTIONS I
l
- Seal Seal injection. Closest approach to Control drive mechanisms Possible damage to the If the scram function Injection pumps to control rod drives wi re trays #109,110,111.
control rod drive on some control rods l System forced circu-and wi re trays.
mechanisms.
Loss of the were lost, the boron.
lation pumps l
wire tray which supplies injection system is j except supply l power to the control rod capable of reactor l
l to control rod drives indicator, etc.
shutdown with all rods l
l dri ves, nonfunctional.
l_.
I I
I l
l Seal l Seal injection At inlet / outlet to Control drive mechanisms Possible damage to the If the scram function l Injection l pumps to filter / accumulator wi re trays #109,110,111.
control rod drive on some control rods l System l forced circu-l section.
I nechanisms.
Loss of the were lost, the boron lation pumps wire tray which supplies injection system is i
except supply power to the control rod capable of reactor to control rod drives indicator, etc.
shutdown with all rods drives.
nonfunctional.
I I
l l Decay Forced Terminus at forced Feedwater line forced None to the feedwater The forced circulation heat circulation circulation loop.
circulation loop valve line due to piping size.
suction and discharge loop 1A to operated by the hydraulic Loss of the hydraulic valves remain in the decay heat valve accumulator system.
valve accumulator control
" as is" position, pump suction.
to forced circulation This will not affect j
- valves, reactor shutdown or
{
emergency core cooling systems.
i Page 9 i
TABLE 2 l
l l SAFETY-RELATED EQUIPMENT l
EFFECT OF BREAK ON REDUNDANT SYSTEMS OR I t l SYSTEM PIPING RUN POSTULATED BREAK IN VICINITY l
SAFETY-RELATED EQUIPMENT MITIGATING ACTIONS I
i Decay Forced Terminus at wall Main Steam Isolation Main Potential Steam Iso-Loss of the shutdown Heat ci rculation into forced Valve, shutdown condenser lation valve operator condenser heat removal Loop 1A to circulation condensate line, failure would leave valve capacity (not required decay heat cubicle.
in "as is" position and for this event) would 4
pump suction, cause scram.
Break of be backed-up by the the shutdown condenser high pressure core condensate lines.
spray system and alternate core spray l
system. The main steam isolation valve is backed-up by the Turbine Building i
Isolation Valve.
I i
i l Decay Forced Suction of the Main stean isolation Main Potential Steam Iso-Loss of the shutdown i
i Heat
. circulation decay heat pump.
valve, wire tray carrying lation valve operator condenser heat removal control rod drive indica-failure would leave valve capacity (not required j
i Loop 1A to l
l l decay heat tions, shutdown condenser in "as is" position and for this event) would pump suction.
condensate line, cause scram. Break of be backed-up by the the shutdown condenser high pressure core l
condensate lines. Loss spray system and l
of control rod drive alternate core spray i
position indication.
system. The main l
steam isolation valve is backed-up by the Turbine Building Isolation Valve.
1 Control rod drive a
position indication l
loss is backed-up nuclear instrumenta-l tion channels 1-8.
1 1
L I
i l
l WP-1 4
e
.c
Page 10 TABLE 2 i
j l
l SAFETY-RELATED EQUIPMENT I
EFFECT OF BREAK ON REDUNDANT SYSTEMS OR SYSTEM l PIPING RUN POSTULATED BREAK i IN VICINITY SAFETY-RELATED EQUIPMENT MITIGATING ACTIONS I
I juecay l Discharge of Tenainus at decay Control rod drive hydrau-Potential Main Steam Iso-Loss of the shutdown l Heat decay heat heat pump or lic line, two wire trays lation valve operator condenser heat removal l
l pump to 1A cubicle or various and main steam isolation failure would leave valve capacity (not required l
l Forced Circu-points between.
valve operator.
in "as is" position and for this event) would l
lation Loop.
cause scram. Break of be backed-up by the j
the shutdown condenser high pressure core l
condensate lines.
Loss spray system and of control rod drive alternate core spray position indication, system. The main steam isolation valve l
is backed-up by the l
Turbine Building Isolation Valve.
Control rod drive l
position indication l
loss is backed-up l
nuclear instrumenta-l tion channels 1-8.
I
\\
l l
l Main Reactor vessel l Anywhere in the Reactor vessel high Possible fracture of the The alternate core l Steam j to containment reactor lower pressure core spray inlet.
high pressure core spray spray system remains l shielding cavity.
line.
available, j wall.
1 Main Reactor vessel Point of leaving Control rod drive Possible damage to If the scram function Steam to containment lower reactor mechanisms.
control rod drive on sone control rods l shielding cavity.
mechanisms.
were lost, the boron l wall.
injection system is j
capable of reactor l
shutdown with all rods l
nonfunctional.
Page 11 TABLE 2 a
l l
l SAFETY-RELATED EQUIPMENT l
EFFECT OF BREAK ON REDUNDANT SYSTEMS OR l SYSTEM l PIPII4G RUN l
POSTULATED BREAK IN VICINITY I
SAFETY-RELATED EQUIPMENT MITIGATING ACTIONS 1
l l Main Reactor vessel Midpoint on Shutdown condenser Loss of shutdown Loss of the shutdown '
l Steam l to containment, horizontal run to condensate line.
condenser.
condenser heat removal l
l shielding Contai ratient capability (not j
j wall.
l Building wall.
required for this l
l event) is backed-up l
by the high pressure l
core spray system.
l l
An analysis of the l
l effects of a double l
l LOCA above and below the core (steam line and shutdown j
condenser condensate l
lines) indicate a j
total flow rate less l
than the design basis l
LOCA.
I l
i i
[Feedwaterl Reactor Containment wall.
None None required.
l l containment vessel wall to forced i
j circulation l
l loop.
l l
l I
l i
lFeedwaterl Reactor Entrance to forced Shutdown condenser Loss of shutdown Loss of the shutdown l
l containment circulation pump condensate lines, condenser.
condenser heat removal i vessel wall cubicle.
capacity (not required to forced for this event) is circulation backed-up by the high loop.
pressure core spray system.
WP-1
Page 12
, TABLE 2 l
1 l
SAFETY-RELATEU EQUIPMENT j
EFFECT OF BREAK ON REDUNDANT SYSTEMS OR l SYSTEM l PIPING RUN l
POSTULATED BREAK IN VICINITY l SAFETY-RELATED EQUIPMENT MITIGATING ACTIONS I
I l
- Feedwaterj Reactor j Piping in forced None NA NA l containment l circulation punp l vessel wall j cubicle.
l l to forced l
l circulation l
l loop.
I i
l _ __ _
1 l
l l
l Feedwaterl Reactor 1ower reactor Boron injection line.
Possible loss of boron The control rod drive l containment cavity.
injection system.
systen is available j
j vessel wall for reactor shutdown.
l to forced l
circulation l
loop.
l l
l l
1 Forced Entire system Any break in lower Boron injection line.
Possible loss of boron The control rod drive lCircula-l contained in reactor cavity.
injection systen.
system is available l tion lower reactor for reactor shutdown. '
l System cavity and two l
l forced circu-l l lation pump l
j cubicles.
l l
l 1
l Forced Entire system l Any break in either None NA NA lCircula-contained in forced circulation l tion lower reactor pump cubicle.
jSystem j cavity and two l
l forced circu-l lation pump l
cubicles.
I I
I I
l l
1 lPurifica-j Reactor vessel j Any break in either Boron injection systen None NA l tion j to biological l forced circulation piping.
jSystem l shield, pump cubicle.
l l
Page 13 TABLE 2 O
l l
l
. SAFETY-RELATED EQUIPMENT I
EFFECT OF BREAK ON REDUNDANT SYSTEMS OR
/
l SYSTEM l PIPING PUN l
POSTULATED BREAK l IN VICINITY l
SAFETY-RELATED EQUIPMENT MITIGATING ACTIONS l
lPurifica-l Heat Exchanger Any break in Variable leg on reactor Loss of reactor water This occurrence would jtion l Cubicle cubicle.
water level safety level indication.
result in reactor l System l
Channels 1, 2 and 3.
l l
This LOCA would l
require reactor con-l tainment building l
l l
flooding to core mic-l l
l plan.
l I
I t
i I
I l Pu rifi ca-l lon Exchanger l Any break in None NA NA l tion l Cubicle l cubicle.
System l
l l
l 1
l l
j Pu ri fi ca-l Pump / filter l Any break in Boron injection piping None NA
[ tion cubicle cubicle l System I
I I
i 1
l Alternate l Reactor vessel At check valve.
High pressure core spray None - Restrained Restrained l Core to check valve l pump suction from over l Spray 38-26-001.
I head storage tank.
l l
I I
I Alternate Reactor vessel Midpoint of hori-2 of 3 main steam relief None - Restrained Restrained Core to check valve zontal run above valves.
l Spray 38-26-001.
l main steam relief l
l l valves.
l l
l
\\
\\
l l
l Alternate: Biological Entry to High pressure core spray None - Restrained Restrained l Core shield to biological shield, pump suction from overhead Spray check valve storage tank.
High l 38-26-001.
pressure core spray system l
valves 53-25-001 and l
l 53-25-003.
1 I
I I
WP-1
Page 14 g
TABLE 2 l
l SAFE TY-P. ELATED EQUIPMENT I
EFFECT OF BREAK ON REDUNDANT SYSTEMS OR f
SYSTEM l PIPING RUN l
POSTULATED BREAK IN VICINITY SAFETY-RELATED EQUIPMENT MITIGATING ACTIONS i
l l
l Alternatej Reactor vessel Inside upper cavityi Reactor vessel.
, None NA l Core to check either at reactor l Spray valve vessel or at cavity l l
l #38-26-001.
l wa11.
l l
l l Control l Collection Either terminus Control rod drives.
Possible damage to If the scram functioq l Rod l header to or the closest control rod drive on some control rods lEftluent i pumps.
approach to the mechanisms.
I were lost, the boron j
, Discharge to l control rod drive injection system is l
l decay heat l mechanism, capable of reactor l
l system.
l shutdown with all l
l control rods l
l nonfunctional.
l l
l l
l l
1 I
I I
i i
l i
l 1
l l
l l
l l
l l
l 1
1 1
1 I
I I
I I
I I
I i
I i
I I
i l
I i
I WP-1
g f l
5
~
~U CVERl-4E AD STORAC,E TA M K.
l 53-25-008
-DEMnN.
S3-25-002 HO t 25-o03
[60-2C-ooco 60-25-002.
l l
/
RC AC.To R.
I VESSEL 30-ns-0o1 l
l l
ScDium PEuTA-
[h
(~ T BoR.ATC ThlK.
=
Go-i s-co l m
O
[lA 16 I53-06 col f3 06-co2
)L ss FRoM RX' FOPLE.D C. n R.C.. SWTEM Morc sesvy L.nacs zuoiCAre co - 2.s-ooi t3cuMDPJES OF HIGH PF'ESSLlRE COR.C S PP AY DISCH AftC,C AMD G o -Z.5 - o o S BOROM.DJJECTION PIPIMG.
C_ceE SeeAv Auo Boron Iuaect O M E.
LIME D I AGR.AK 03,,4 jgy,y
- w. o.
w o.
se a,Ly DRG.NO.
DAIRYLAND POWER COOPER ATIVE A/D LA cnossa, wescoNSIN DATE J/25/62 i,,. o
- _ _ _ _ _ _ _ _ _ _ _ _ _ _