ML20042C097
| ML20042C097 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/26/1982 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | Miraglia F Office of Nuclear Reactor Regulation |
| References | |
| SBN-243, NUDOCS 8203300194 | |
| Download: ML20042C097 (18) | |
Text
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SF#KOK STATION Engineering Office:
1671 Worcester Road Framingham, MA 01701 March 26, 1982 9
SBN-243 E U C 1 'E O I
T.F.
B7.1.2 9
MAR P,919929-S United States Nuclear Regulatory Commission
~ C FCM TI":T"'wV"::3 Washington, D.C.
20555
//
Attention:
Mr. Frank J. Miraglia, Chief gy Licensing Branch No. 3 Division of Licensing
References:
(a)
Construction Permits CPPR-135 and CPPR-136 Docket Nos. 50-443 and 50-444 (b)
USNRC Letter, dated February 12, 1982, " Request for Additional L. formation," F.J. Miraglia to W.C. Tallman (c)
PSNH Letter, dated March 12, 1982, " Response to 440 Series RAIs (Reactor Systems Branch)," J. DeVincentis to F.J. Miraglia
Subject:
Response to RAI 440.61; (Reactor Systems Branch)
Dear Sir:
We have enclosed the response to the subject RAI, which you forwarded in Reference (b):
This response was not included in the Reference (c) responses.
Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY in
/&
ohn DeVincentis Project Manager Attachments j
0 6'(\\
8203300194 820326 PDR ADOCK 05000443 A
440.61 Provide a discussion of long-term ef fects and events for each accident analyzed in Chapt er 15, assuming no operator action prior to times justified by ANSI-M660.
When operator action is needed, provide a complete assessment of the operator's role and show that sufficient time is allowed for operator action to be acen plished.
Verify that the acceptance criteria for each accident a re not exceeded in the long-term.
RESPONSE
For most of the events analyzed in Chapter 15, the plant will be in a safe and stable hot standby condition following the automatic actuation of reactor trip.
This condition will in fact be similar to plant conditions following any normal, orderly shutdown of the reac t o r.
At this point, the actions taken by the operator would be no dif ferent than normal operating procedures. The exact actions taken, and the time these actions would occur, will depend on what systems are available (e.g., steam dump system, main feedwater system, etc.) and the plans for further plant operation. As a minimum, to maintain the hot stabilized condition, decay heat must be removed via the steam generators.
The main feedwater system and the steam dump or atmocpheric relief system could be used for this purpose.
Alternatively, the emergency feedwater system and the steam generator safety valves may be used, both of which are safety grade systems. Although the emergency feed system may be started manually, it will be automatically actuated if needed by one of the signals shown on Figure 7.2-1, sheet 15, such as low-low steam generator water level.
If hot standby conditions are maintained for an extended period of time, operator action may be required to transfer the emergency feedwater source. The time at which such action is required will be sufficiently long after initiation of the event to permit operator action. Also, if the hot standby condition is maintained for an extended period of time (greater than approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />), operator action may be required to add boric acid via the CVCS to compensate for xenon decaz and. maintain shutdown margin. Again, theactionstckenbYthIoperatorwould be no different than during normal plant shutdown.
Many Chapter 15 events result in a stable condition being reached automatically following a reactor trip and only actions typical or normal operation are required from the operator. For several events involving breaks in the reactoc coolant system or secondary system piping, additional requirements for operator action can be identified.
( Additional information about the impact of equipment failvi es or erroneous operator acticus may be found in WCAP-9691 "NUREC-0578 2.1.9.C, Transient and Ac2ident Analysis".)
Steamline Break:
See Table 440.61-1 Following the hypothetical steamline break incident, a steamline isolation signal will be generated almost immediately, causing the main steam isolation valves to close within a few seconds.
If the break is downstream of the isolation valves, all of which subsequently close, the break will be isolated.
If the break is upstream of the isolation valves, or if one valve fails to close, the break will be isolated to three steam generators while the
\\_
faulted steam generator will continue to blow down.
Only the case in which the steam generator continues to blowdown is discussed here since the downstream break followed by isolation of all steam generators will terminate the transient.
An excessive cooldown protection signal will cause main feedwater isolation to occur.
The only source of water available to the f aulted steam generator is then the emergency feedwater system.
The first required operator action is to identify the faulted steam generator and verify that automatic isolation of emergency feedwater flow to the faulted steam generator has taken place.
Automatic isolation of EFW flow to a f aulted steam generator is accomplished by flow sensing devices which sense an abnormally high EFW flow to one of the four steam generators and automatically close the isolation valve to that steam generator.
In the case of smaller steamline breaks where an abnormally high EFW flow is not created, manual isolation capability exists once the operator has determined the faulted steam generator.
Following steamline isolation, steam pressure in the steamline with the faulted steam generator will continue to fall rapidly, while the pressure stabilizes in the remaining three steamlines.
The indication of the different steam pressures will be available to the operator within a few seconds of the steamline isolation.
Additionally, EFW flow indication to each of the four steam generators is provided on the main control board. This will provide the necessary information to identify the faulted steam generator so that emergency feedwater to it can be isolated if automatic isolation has not yet occurred. The operator is instructed by Emergency Operating Procedures to isolate the af fected steam generator by shutting the sceam generator EFW isolation valve. Manual controls are provided in the control room for start and stop of the EFW pumps and for the control of isolation valves associated with the EFW system. The means for detecting the faulted steam generator and igol,ating emergency 2,2 feedwater to it requires only the use of safety grade equipment available following the break.
Following the automatic safety injection actuation and af ter the faulted steam generator is completely isolated,sthe continued operation of the safety injection system will repressurize the reactor coolant system and continue to_ increase the RCS volume i nvento ry. The second required operator action is to manually control the repressurization of the reactor coolant system and modulate safety injection pumps to control pressurizer level. The operator may then restore normal pressure and level control and stop the safety injection pumps.
The operator has available, in the control room, an indication of pressurizer level from the instrumentation in the reactor protection system.
To maintain the indicated water level, the operator can start and stop the centrifugal charging pumps as necessary. As soon as an indicated water level returns to the pressurizer, RCS pressure has returned to the normal range and the RCS is sufficiently subcooled, the operator is instructed to stop the safety injection pumps, re-establish normal charging and letdown flows, and re-establish
operation of the pressurizer heaters to maintain a steam bubble in the pressurizer to limit system repressurization.
The pressurizer level instrumentation and manual cont rols for operation of the high head SI pumbs meet the required standards for saf ety systems.
The removal of decay heat in the long-term (following the initial cooldown) using thc remaining intact steam generators requires only the emergency feedwater system as a water source and the secondary system safety valves to relieve steam.
The requirements to terminate emergency feedwater flow to the faulted steam generator, re-establishing normal charging and letdown flows, and re-establishing operation of the pressurizer heaters can be met by simple switch actions by the operator.
Thus, the required actions to limit the cooldown and repressurization can be easily recognized, planned and performed within ten minutes. For decay heat removal and plant cooldown the operator has a considerably longer time period in which to respond because of the large initial cooldown associated with a steamline break transient.
Feedwater Line Break: See Table 440.61-2 For a feedwater line break, emergency feedwater is initiated automatically, as is safety injection.
For the feedline break downstream of the main feedwater isolation valves, the required operator actions are similar in nature to the required actions for the steamline break.
The first required operator action is to identify the faulted steam generator and verify that automatic isolation of EFW flow to that steam generator has taken place. The primary indication to the operator will be a comparison of individual steamline pressures after steamline isolation has occurred and EFW flow indication to each steam generator. Af ter identifying the faulted steam generator, the operator is instructed to isolate EFW flow to that steam generator by shutting the steam generator EFW isolation valve if automatic isolation has not yet occurred. The steamline pressure indicators, EFW flow indicators and EFW isolation valves are safety grade.
The operator must provide for decay heat removal through the intact steam generators by maintaining steam generator water level using emergency feedwater as a makeup supply. The operator can use the steam dump system or the steam generator ARV's to begin a controlled cooldown, or the unit may be maintained in hot standby by using the steam side safety valves for decay heat removal.
Finally, the operator must modulate the high head safety injection pumps to control primary pressure and pressurizer level. The operator must observe the primary steam pressure-temperature relationship to ensure that voiding does not occur in the reactor coolant system. The operator uses safety grade instrumentation and controls to manually control the primary system pressure and maintain normal pressurizer level.
The analysis presented in FSAR Section 15.2.8 assumes a 30-minute de I..f.nti1 tbese actions occur.
Boron Dilution (later)
Steam Generator Tube Rupture:
See Table 440.61-3 The accident examined is the complete severance of a singic steam The operator is expected to determine that a
generator tube.
steam generator tube rupture has occurred, and to identify and isolate the faulty steam generator on a restricted time scale in order to minimize contamination of the secondary system and ensure termination of radioactive release to the atmosphere f rom the The recovery procedure can be carried out on a time faulty unit.
scale which ensures that break flow to the secondary system is terminated before water level in the affected steam generator rises into the main steamline. Sufficient indications andthese controls are provided to enable the operator to carry out functions satisfactorily. Consideration of the indications provided at the control board, together with the magnitude of the break flow, Icads to the conclusion that the isolation procedure can be completed within 30 minutes of accident initiation.
Included in this 30-minute time period would be an allowance of 5 minutes to trip the reactor and actuate the safety injection system (automatic actions),10 minutes to identify the accident as a steam generator tube rupture and 15 minutes to isolate the faulted steam generator.
Immediately apparent symptoms of a tube rupture accident such as falling pressurizer pressure and level and increased charging pump flow are also symptoms of small steamline breaks and loss of coolant accidents. It is therefore important for the operator to determine that the accident is a rupture of a steam generator tube in order that he may carry out the correct recovery procedure.
The accident under discussion can be identified by the following method:
in the event of a complete tube rupture, it will be clear soon after the trip that the level in one steam generator is rising more rapidly than in the others.
Also, this accident could be identified by either a condenser vacuum pump exhaust high radiation alarm or a steam generator bloydown radiation alarm.
The operator carries out the following major operator actions subsequent to reactor trip which lead to isolation of the faulted steam generator and minimizing primary to secondary leakage:
1.
Identification of the faulted steam generator.
2.
Isolation of the faulted steam generator.
3.
Subcooling of RCS fluid to 500 below no-load temperature.
I I
4.
Ih pressurization of the RCS to t e rm i na t e breakflow, and 5.
Terminating safety injection.
SulIicient indications and controls are provided to enable the operator to complete these functions satisfactorily.
Table 440.61-3 lists applicabic instrumentation and equipment, their associated safety grade classifications, and the impact of a single active component failure.
Loss of Coolant Accident: See Table 440.61-4 No manual actions are required of the operator for proper operation of the ECCS during the injection mode of operation.
Only limited manual actions are required by the operator to realign the system for the cold leg recirculation mode of operation, and, at approximately 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />, for the hot leg recirculation mode of operation.
These actions are delineated in Table 440.61-4.
The changeover from the injection mode to recirculation mode is initiated automatically and completed manually by operator action from the control room. Protection logic is provided to automatically open the two safety injection system (SIS) recirculation sump isolation valves when two of four refueling water storage tank level channels indicate a refueling water storage tank level less than a low level setpoint in conjunction with the initiation of the engineered safeguards actuation signal
("S" signal). This automatic action would align the two residual heat removal pumps to take suction from the containment sump and to deliver directly to the RCS.
It should be noted that the residual heat removal pumps would continue to operate during this changeover from injection mode to recirculation mode.
The two charging pumps and the two safety injection pumps would continue to take suction from the refueling water storage tank, following the above automatic action, until nan tal operator action is taken to align these pumps in series with the residual heat removal pumps.
The refueling water storage tank low level protection logic consists of four level channels with each level chant.el assigned to a separate process control protection set.
Four refueling water storage tank level transmitters provide level signals to corresponding normally deenergized level channel bistables.
Each level channel bistable would be energized on receipt of a refueling water storage tank level signal less than the low level setpoint.
A two out of four coincident logic is utilized in both protection cabinets A and B to ensure a trip signal in the event that two of the four level channel bistables are energized. This trip signal, in conjunction with the "S' signal, provides the actuation signal to automatically open the corresponding containment sump isolation valves.
_ _ =.
The lesw ref ueling water storage tank Icvel signal is a lms a la rmed to inform the ope ra'_o r t o initiate the manual action required to realign the charging and safety injection pumps f or the recirculat ion mode.
The manual switchover sequence that must be perf ormed by the operator is delineated in Table 440.fil-4.
Following the automatic and manual switchover sequence, the two residual heat removal pumps would take suction from the r
containment sump and deliver borated water directly to the RCS cold legs.
A portion of the number 1 residual heat removal pump i
discharge flow would be used to provide suction to the two charging pumps which also deliver directly to the RCS cold legs.
A portion of the discharge flow from the number 2 residual heat removal pump would be used to provide suction to the two safety injection pumps which would also deliver directly to the RCS cold legs.
As part of the manual switchover procedure, the suctions of the safety injection and charging pumps are cross connected so that one residual heat removal pump can deliver flow to the RCS and both safety injection and charging pumps, in the event of the failure of the second residual heat removal pump.
I 4
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L s
I n
TA31.E 440.61-1 STEAMLINE 3REAK Safety Crade impact of Operator's Classification of the impact of Failure to Take Instructions Components and Single Action or the Reqrirsd Alarus to Alert the Civen to the Operator Instrumentation Active Operator Taking a Operater Operator to Initiate Delay Time for Performing the Necessary to Complete Component Closely Related Action A Particular Action Assumed Required Action InJ1cated Action Failure but Erroneous Action A.
Id:ntify the A.
Primary indication to A.
Within 10 A.
Identify the faulted A.1 Steam line pressure None A.
The operator does fsaltid steam the operator is steamline minutes steam generator by indicators.
not isolate EFW to ginaritor and pressure todication and comparing steamline any steam generater iso 11t3 emergency individual EFW flow pressures and individual A.2 Steam generator None or isolates EFW to fudwater to that indication to each steam EFW flow indication to ETW control valves, wrong steam stian generator if generator. A possible each steam generator.
generator The entomatic isolation alare is the steam Terminate emergency A.3 Steam generater None faulted steam hai not yet flow-feed flow mismatch.
feedwater to that steam level indleators.
generator u111 ocesrrid.'
generator by shutting continue to blowdown.
the EFW isolation valves A.4 EFW flow indication.
If automatic isolation (All safety grade.)
has net occurred.
5.
The operator 3.
Primary indications B.
Within 10 B.
The conditions for 8.1 Pressuriser None 3.1 The operator must rsset the to the operator ares minutes resetting safety level indicators.
f ails to modulate s&fsty injection pressuriser level, injection are given to
$1 pumps af ter the and marually pressuriser pressure and the operator. The 3.2 Pre ssuriser None pressuriser level contest the repres-RCS temperature.
operator is instructed pressure indicators.
retusas to the ssriastion of RCS Possible alarms includes to manually control the indicating ranges sad maintain normal
- high pressuriser level high head $1 pumps and 3.3 RCS temperature None Water rettei a
prsssare control.
- high pressuriser re-estab!!ah normal indicators.
through pressuriser pre s su re.
pressuriser level (All safety grade.)
relief valves control.
may occur.
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TAB 12 440.61-1 (Coettened) l STrAPE.1NE SafAK l
,-\\lf.
Safety Crede tapact of Operator's Claselfleettee of the femt of Fs11ere to Take fostructions
- , _ ets and Staale Action er the i
Begv. ired Alerne to Alert the C1,ee to the Operator featrumentation Active operator Taking a Operstor Operator to lettiate.
Delay Time for performing the Necessary to Ceeplete Ceeposest Closely Related Action A porttenter Aettes Assumed Eeguired Action ledicated Action pe11ere but Erroneous Action 3.4 Eigh head safety Nome tejection pumpe.
(All safety grade.)
t I,
3.2 The operator stope
[
$1 before peak reactivity is reached:
i If criticality le I
attelmed, the core l
power u111 leeresse gr.39s j
until it reaches j
equilibrium uith stese demand.
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1 TABLE 440.61-2 FEEDLINE BREAK laetructietw Componeets and lepact of Operator's Reqstr:d Alarme to Alert the
- Civen to the Operator lastrumentation tapact of Failure to Take operstar Operator to f attiste Delay Time for Performing the Necessary to Complete Strale Aettom or the 4
ll Action A Fartteeler Actica Asevned Required Action ladicated Action Active Operator Taking a Componeet Closely Related A.
The operator A.
Primary ladication A.
Withia 30 A.
Ideotify the faulted A.1 Steaaline pressure Failure but Erroneous Action
>l should identify to the operator will j minutes S/C by comparing ladividual ladicatora.
('
the faulted steam be a comparison of i (
oteamline pressures and None A.1 Operator does not i
generf. tor and tae-steaalise pressures ;
individual EFW flow A.2 Steam generator recogolae the accident 1:t3 ETV to that if and individual EN flow sadication to each stese EFW control valves.
and does not isolate Gttomatic isolation todication to each 6
generator. Secure EFV None EN to any steam has ext yet occurred.
steam generator. Possible flow to the faulted S/C by A.3 Steam generator generators Faulted alarus loclude 3
shuttias the EFW isolation level ladications.
steam generator will
- stema/ feed flow mismatch valves for that S/C if None continue to blowdown.
- high E N flow
,)
automatic toolatloa has A.4 EFV flow A.2 Operator isolates a
- low steen generator 3evel not yet occurred.
ladication.
ETV to wrong stese
- low mata steam pressure (All oefety grade.)
generaters
{
f The faulted stema e
e generator will 3.
The operator con-3.
The operator will 3.
Within 30 3.
Maintata proper S/C 3.1 Steam Cenerator continue to blowdown.
trs13 EFW to the use ladividual S/C level minutes level la intact S/C's.
EN valves and controle.
{
1; tact steam gen-ladicattoa to control.EN if posalble, mastaine (Safety grade.)
None 3.1 The operatos falla 4
tritirs and controle flow to each of the' steam EFW flow to intact S/C's to control EN flows sooldown.
generators. Righ level to help lower primary to intact etese and low level alarme j temperature.
generators:
I are provided.
- ?
Overfilling of a steau 4;
generator -, ouur.
i
- - i 4
1 L
t s
4 b'
n i
f TA311440.61-2 (Continued)
FT.im,1NE BEEAK lopect of Operator's pattere to Take Safety Crade impact of Action er the Classification of the
$1mste Operator Taking a Componente and Active Closely Related lastructions lastraneetattos Componeet bet Erroneous Action _
C1,em to the Operator pecessary to Coop 1ste Fatture _
5 for Forforming the ladicated Aetten Alarms to Alert the Delay Time Regelred Action C.1 Operator Fone tamediately resets
,Rectred Operator to Initiate Oper: tor A Ferticular Action _ j Assumed _
C.1 N1 beed $1 pump the si signal and Reset 31 and modelete controle.
C.
J C.
Within 30 high head SI pump flow stops the pumps:
Aetten pressuriser level and stoutes C.
The rperator pressere ladicattoa and high to control primary system Totding to t e reactor C.
h modelttas the high and low level alarse are pressure and presseriter pone coolant system may occur level. cheerve the C.2 pre sserizer head si pumps to provided.
contr:1 primary primary erstes pressure-pressure indicators.
C.2 The operator doen temperatete relationshly
&7;tes pretoure and j
to encore that the SCS le C.3 Pressuriser pressortser level None act modulate $1 after preisertsJr level.
/
sufficiently subcooled.
level indicators.
returns to the None indicaties ranges C.4 RCS ten Hrtature Water rettet through f
ladicators.
the presavelser (All safety grade.)
relief valves may l
occur.
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W.
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'"' b T h.M-Ecachem.M wif.jr f.,
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.- ; y.g 3,.
. A 6..
w
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n TABLE 440.61-3 i
STEAM CENERATOR TU3E RUFTUIE Safety Crede tapact of Cperator's Alarse to Alert the Classification of the lepact of Fatture to Take Operator to initiate lastructione components sad Stagle Action or the Requirst A Particular Action Cives to the Operator lastrumentation Active Operator Taktag a Operrter and Their Safety Delay Time for Perferetag the Necessary to Complete Componest Closely Related Assumed Required Action Indicated Action Falture but Erroneous Action Aetton Crade Class i
- 1 A.
Possible alarme loclude
- A.
30 stoutese A.
The operator is A.1 Stese Cenerator A.
Several diverse A.
Fallere to identify A.
Id ntify faulted
- staae generator level (high lastructed to identify Level Indicatore lastrument ladications the faulted S/C otras generator (S/C).
- steam / feed flow elematch '
f aulted S/C by one er more are Safety Crade (FAMS).
are included as would result la act
- blowdown radletten (high).
of the fellowleg methods:
that any stagle teolating the feelted
- esta stees radlettoe (high) a) high S/C level la A.2 slowdows Line active component S/C which is addressed one S/C; radiatton nomitor fa11ere would in the Westtaghouse b) high radiation free la non-safety grade.
not preclude the Owners Group procedures l
eay one steami operator free Development and i
generator blowdova A.3 Means for detersta-identifyl g the Evaluation Progree for N' REC-073 7, Ites 1.C.I.
}
time radiation nositer 1rts high radiaties faulted S/C.
J
?) high radiation free la one S/C, e.g.,
I,'
any ese stese saepting to songafety generator (e.g.,
grade.
samplies);
d) high radiation free any one stese generator esta steamlise.
3.
30 atoutes*
B.
The operator is 3.1 EFW control 3.1 Mone 3.
Failure to isolate f" I 3.
13 slate S.
No alares necessary lastructed tor valves are safety grade.
Faulted S/C to Fasitto S/C.
addressed la I
(
k e
t s
~
- i TAst.E 440.61-3 STEAtt CENERATOR TUSE R'JFTURE (Coattaved) tapact of Operator's Failure to Take Safety Grade 18 Pact of Classificattom of the Actica or the Slaste operator Taking a Components and Active Alares to Alert the Instrections Instrueestation Closely Related Component Operator to Initiate Civen to the Operator Necessary to Complete but Erroneous Action for Performing the
!=dicated Action Fatture A Ferticolar Action Delay Time Regstrid and Their Safety Assumed
_Dequired Action the Westinghouse Operatir Crede Claes B.2 ftSIT's are safety 3.2 pone Owners Croup Frocedure
_ Action a) stop all feedwater Development and flow to faulted steam grade.
generator:
s.3 Steam gener. tor 3.3 Discussed Evaluation Program for NUSEC-0737 b) close the main aream ARVs are non-saf ety la WCAF-9691 for Ites 1.C.1 1solettoa valve and event tree 1
bypass valves grade.
sequences associated essociated with the Isolattom valve (s) with secondary side with the faulted S/C; 3.4 relief valves c) verify closure of all to steam drives EFW Is11ere to opes and atmospheric rettel pumps are safety grsde.
reclose.
valves associated with the faulted S/C; d) close the isolatica valve la the stese itse to the emergency feedveter pump sesociated with the if faulted S/C; C.
Fatter, to cool down C.1 None C.
30 stoutes* ne operator is instructed c.1 Stesettne pressure RCS via non-faulted ladicators are safety S/C is addressed in C.
Cal down RCS to Primary indicatione to rapidly cooldown the Westinghouse are steseline pressure the RCS to 50*F below grade.
C.2 None C.2 Reactor coolant owners Croup 50*F beltw no-and SCS temperature.
no-load temperature temperature indicators Proeedure Development load temperature by erese dump from are saf ety gree.
and Evaluetten Progree by ese of steam non-faulted S/C to con-Steam dump esoldown for NUREC-0737, Ites I.C.I.
C.3 dump.
denser.,8f off-site power valves are non-safety and condenser are avattable, or by opening grade.
S/C ARV's.
$s
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4 se--
{
n
-r-A A,y
i TABLE 440.61-3 STEAM CENERATOR TURE RUFTURE (Continued)
Ala rna to Alert the Safety Crade Impact of Operator's Claselfication of the Impact of Failure to Take Operator to Initiate Instructions Componente and Single Actice or the Rettirro A Farticular Action Civen to the Operator Instrumentation Active Operator Taking a Operstir and Their Saf 3ty Delay Time for Performing the Necessary to Complete Component Closely Related Aceton Crede Class Assumed Required Action Indicated Action Failure but Erroneous Action
' I
!! off-site power or C.4 Steam generator
{
condenser are not ARV's are non-safety available.
grade.
D.
Depressertse RCS Primary indicationg D.
30 etautes*
D.
The operator le D.1 Normal pressurtser D.1 Pressurtser spray D.
Failure to depressurt se I
ts faulted S/C and SC and RCS pressure.
teatructed to depressurise spra) le non-eafety to provided by 2 out of RCS to addressed in pressors.
RCS to faulted S/C pressure grade.
the four RCP's. With the Westinghouse by utilising the following loss of off-stte Ownero Croup Procedure methods:
D.2 Pressuriser PORV's power RCF's will not Development and a) if RCP's in are safety grade.
be available to provide Evaluation Program f or service, using normal normal spray.
NUREC-0737. Ites I.C.1 i
pressuriser spray; D.2 Two pressuriser I
b) If RCF's are not PORV's are provided in la service, using the pressertner a nd pressuriser Port; one to sufficient c) If above 2 methode to depressurise RCS.
are unavailable, uslag semillary spray; D.3 Aux 111ery spray le non-aafety grade.
D.3 Impact of losing all means of i
depressuristre RCS is addressed la the
.W i
h
t a
I
\\
I TASLE 440.61-3 l
STEMI CENERATOR TURE RUPTURE (Continued)
Safety Crade Class-Impact of operator's Alares to Alert the ification of the 1spect of Pa11ure to Take Operator to initiate
- Instructions Components and Single Action or the Regit t'ed A Particular Action Civen to the Operator Instrumentation Active Operator Taking a Operatar and Their Safety Delay Time for Performing the Necessary to Complete Component Closely Related Actise Crede Class Assumed Regelred Action Indfeated Action Pa11ure but Erroneous Action Westlighouse Owner's Croup Procedures Development and Evaluation Program i
for NURIC-0737, Iten 1.C.I.
h E.
7;rsinate St.
E.
Primary indleations E.
30 minutes
- E.
The operator le E.1 Safety injection E.1 None E.
The f ailure to terminate are pressuriser level, uhen RCS pressure increased toetructed to terminate S1 system to safety grade.
51 is addressed in the RCS pressure, and RCS l
Westinghouse Owner's temperature
- {
by 200 pet, there le en E.2 pressuriser level E.2 None Group Procedures indicated pressortaer indicatione are Development and Evaluation level, and RCS subcooling safety grade (PAMS).
Program for NUREC-0737, le verified.
Item I.C.1 E.3 Reactor coolant E.3 Mone systen pressure indications are safety grade (PAMS).
E.4 'Aeactor coolant E.4 71one gv system temperature indleations are safety grade (PAMS).
- The above 5 steps are sequential and it le assumed that for felt double-ended steam generator tube rupture that all 5 allt be completed in 30 minutes af ter initiation of the event.
TABLE 440.61-4 LOSS OF COOLANT ACCIDENT Instructions Given Impact of the Operator's Required to the Operator for Failure to Take Action or the Operator Performing the Operator Taking A Closely Ac t i on Required Action Related but Erroneous Action A.
The operator
- 1) Verify that the con-A.6B.
The plant emergency must manually tainment sump isola-operating procedures complete the tion valves are open.
include instructions and changeover of
- 2) Close the isolation verification steps to the ECCS sys-valve in each RilR ensure proper manual tem from the suction line from the realignment of the ECCS for recirculation by the injection mode RWST.
to the cold
- 3) Close the two isola-operator. The failure leg recircu-tion valves in the to perform one step or lation mode.
crossover line down-the performance of one stream of the RilR heat step out of order, as a exchangers single failure, should
- 4) Close the isolation not reduce ECCS recir-valve in each SI pump culation capability miniflow line.
below minimum safe-
- 5) Open the valve in the guards. Should the discharge line from the operator fail to take number 1 RRR heat any action following exchanger to the suction automatic ECCS switch-of the centrifugal over initiation, the charging pumps; open the consequences will be the valve in the discharge loss of the safety line from the number 2 injection and charging RilR heat exchanger to pumps. The residual the suction of the SI heat removal pumps will be protected from damage by pumps.
- 6) Open the two parallel automatic ECCS switch-valves in the common over initiation.
For
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suction <line between a small break LOCA in the the centrifugal charg-unlikely event of losing ing pump suction and all high head pump the SI pump suction.
delivery capability
- 7) Close the two parallel this situation could isolation valves in the lead to core uncovery centrifugal charging and iradequate core pump suction line from cooli.g. This situation the RWST; close the is addressed in WCAP-isolation valve in the 9691 as the loss of the SI pump-suction line emergency coolant recir-from the RWST.
culation (ECR) function following a small break B.
At approxi-
- 1) Insure that the "S" LOCA. Analyses have mately 17 signals have been reset been performed for loss hours af ter and defeated if not of high head safety the transient previously accomplished.
injection for small LOCA l
TA!!LE 440.61-4 (Cont i nued )
LOSS OF COOLANT ACC It'l:NT Instructions Given Impact of the Operator's Required to the Operator f or Failut* to Take Action or the Operator performing the Operator Taking A Closely Ac t ion Required Action Related but Erroneous Action,
is initiated,
Inadequate must manually establish RilR pump flow core cooling guidelines switch over to to the RCS hot legs by:
are addressed in the the hot leg a) closing the RIIR cold Westinghouse Owners recirculation leg header isolation Group Procedures Devel-ment and Evaluation Pro-mode.
valves; b) opening the two iso-gram for NUREG-0737 lation valves in the Item I.C l.
crossover line down-stream of the RilR For large break LOCA the heat exchangers; residual heat removal c) opening the RilR hot pump delivery to the RCS leg header isolation would be sufficient to valve; provide adequate core cooling during recir-culation.
a) stop SI pump no. I and close its cor-responding cold Icg crossover header isolation valve; b) open its corre-sponding hot _ leg header isolation valve; c) restart SI pump no. 1; d) stop SI pump no. 2 and close its cor-responding cold leg crossover header isolation valve; e) close the SI common cold leg header iso-lation valve; f) open SI pump no. 2's hot leg header iso-lation valve; l
I u.
TAllLE 440.61-4 ( U)n t i nimd )
LOSS OF COOL. ANT ALJ1Dl:.NT Instructions Given Impact of t he Ope ra t o r 's Required to the Operator for Failure to Take Ac t ion or the Operator Pe rf orming t he Operator Taking A Closely Action Requ ired Ac t ion Ielated but Erroneous Action g) restart SI pump no. 2.
C.
Check reactor C.
The operator is instructed C.
Di-cossed in WCAP-9584.
coolant pump to trip all RCP's when the trip criteria.
RCS pressure reaches a specified pressure and SI operation is verified.
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