A08995, Rev 1 to Auxiliary Initiation Event Analysis
| ML20059M187 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 09/14/1990 |
| From: | Scace D NORTHEAST UTILITIES |
| To: | |
| Shared Package | |
| ML20059M186 | List: |
| References | |
| A08995, A8995, NUDOCS 9010030258 | |
| Download: ML20059M187 (16) | |
Text
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Docket No. 50-213 A08995 l
.c Attachment No. 1~
i Haddam Neck Plant-1 AFW Initiation Event Analysis
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l, 9010030258 90o919 fDR ADock 0500021 PDC September 1990
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- c AFW INITIATION EVENT ANALYSIS
-Rev. 1 Performed by.
& o m.'f / 4.~ -9A Wro Daniel.R. Scace/ Date Senior-Instructor NUSCO Operator Training, Branch Reviewed by-Nf YN 90 Jgre IjiPla
/ D&t6 Operations Man er,_CYAPCO
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3 PURPOSE
- 1. This document provides an analysis of an auxiliary feed water
( AFW)-
initiation event: with best-estimate assumptions using ANSI /ANS 58.8-1984, Time Response: Design; l
Criteria for Nuclear Safety Related Operator Actions. Two scenarios are analyzed. The first involves manual (Case - 1)just operator action in the. control room -to ad AFW flow to-greater than, 320 gpm. The second (Case 2) requires local manual action to start an AFW pump to supply AFW flow.
2.
Also provided are times for the same two scenarios L for -
completing the operator action - required to - sup
- flow, by walk-throuch of the required operator action. ply AFW:
The analysis described in 1.
above shows that case 1 meets the requirements of ANSI /ANS 58.8 for. crediting operator action from a best estimate analysis standpoint. The case 2:
analysis is provided' for information only for: this low probability event. The plant' walk-through of Case 1, in which.
the operator takes action in the control room,c shows that the required action can be accomplished within a design basis time of 4 minutes.
The-Case 2 walk-through is provided for information only because there.is no. design basis-for.
comparison for this event.
EVENT DESCRIPTION - CASP 1 This event is initiated by both main feed water pum
- tripping, which will' also cause automatic initiation ' ps of auxiliary feed water. Feef " low will decrease causing steam generator level to dacrease. A reactor trip occurs when the low level setpoint of 11% narrow range'is reached. At the reactor trip the operator will enter E-0, complete E-0 to step 4, and most likely proceed to FR-H.1,LResponse to Loss of Secondary Heat Sink, due to a Red Path existing on Heat Sink.
This condition exists due to wide range. steam generator levels being less than 63% and AFW flow less than 320 gpm. AFW flow requirements are not met. in 'this scenario because only one pump is operating and automatic initiation deficiencies prevent design flow from being achieved. If steam generator level was greater than 63% the red path would not exist and the operator would proceed to-ES-0.1. The FR-H.1 flowpath is more likely and more limiting due to the amount of notes in the beginning of the procedure, and is therefore used for this analysis.
The.320 gpm flow requirement is achieved by the control-room operator manually adjusting steam to the AFW pump turbine at step 2 of FR-H.1 or ES-0.1.
Adjusting AFW flow requires the operator to complete four manipulations; defeat AFW, reset-AFW WL, start an AFW pump from the main control board and open steam a
generator feed water bypass valves.
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EVENT DESCRIPTION - CASE 2
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This event is assumed to be initiated by a seismic event which causes loss, of offsite power, which causes automatic initiation of.AFW, and loss'of' control air.
A reactor trip occurs within 2 seconds.-
Loss of control - air. causes the rapidly, causing the -Terry turbines y. turbines toto-trip on-overspeed. On steam admission valves on the Terr open.Very automatic initiation of AFW,.an operator is dispatched to'.the Terry turbine room.
There the operator is. required to-perform 5
manipulations to-restore AFW flows close the auxiliary - feed-pump. steam supply control. valve, resets the auxiliary feed pump overspeed ' trip, close the air. supply:
. isolation valve to the steam supply control valve, open the air - regulator drain valve (not required in this scenario since the-air lines are already depressurized, but required by the procedure), and. start the AFW pump ~by opening the-steam supply control valve.
The-operator will open the valve as required' to establish a minimum of 320. gpm - based on communication with the control room.. Procedure flow path in a
this scenario would be as. discussed in case.1 or directly to ECA 0.0, Station Blackout, where AFW flow is addressed in step 4.
The flow path is not ' critical to timing of this i
- event, however, because the operator. going to the Terry-turbine room is the critical portion-of the scenario'and is-dispatched when automatic AFW' initiation'is recognized by the control room operators.
ANSI /ANS 58.8 ANALYSIS - CASE 1 Timing for this analysis is defined below forithe specific event based on criteria in ANSI /ANS 58.8.
to = event initiation For this event, the time both main feed pumps-trip, te = event alarm L
The time after event initiation that an alarm occurs to identify the event to the operator. For this event, alarms l
indicating a reactor trip due to low steam ' generator level coincident with SP/FF mismatch identify the time. This time l
is 7
seconds after the. feed pumps
- trip, based on best estimate analysis.
ta = operator action alarm The time after event initiation that the operator-identifies the need - for o this scenario, perator action based on an alarm signal. For there are several indications-to the operator to indicate the need for APW flow.
These include AFW.
2 4
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.t initiation at the time'both mainifeed pumps. trip,-the reactor
- trip on steam-generator low' level coincident with SF/FF mismatch, and ~ the red path which exists on Heat. Sink. For purposes of this analysis the red path is used,-because this.
J is a positive indication,that AFW flow.'is! inadequate and must-be increased Although the red path ' may not, be ' addressed until exiting E-0 at step 4,.the' red path'will exist'at tho' moment _ the-reactor
- trips, or 7
seconds.
after event q
initiation.
tm = time margin complete The earliest time' af ter. te = at= which operator action can be -
considered.
This. is-5 minutes for this ' event, which - is considered to be a Plant condition-2. event.
tp = complete: safety function The time at which the nuclear safety related function must be complete to prevent design criteria from being exceeded. For this event AFW flow must-be established at 320 gpm-15 minutes-after the main feed pumps trip.
1 to = complete operator action The time operator action must be complete to insure d'esign criteria are not exceeded.
For this event, tc tp
=-15
=
1 minutes since the operator-will not-know his. action 'is complete until he has adjusted AFW flow to greater than 320 gpm, which is the design criteria to-be met, ti = latest time to initiate operator-action This number is calculated based on the.
number -of t
manipulations. required to be performed to ~ complete-the.
required operator action. No fixed time.is included since the' operator action required is already identified prior to tm.
?
Since this scenario requires 4 manipulations to increase AFW flow, 4 minutes are required. The ti time is: established by subtracting 4 minutes f rom tp, or 11 minutes after :the main feed pumps trip.
The time line based on this timing criteria is shown in Tigure 1. Note that tm establishes.the time that the operator
- an be credited with taking action. Since ti occurs after tm, ndequate time is available for operator action to restore AFW ilow to greater than 320 gpm.
ANCI/ANS 58.8 ANALYSIS - CASE 2 Timing for this specific event is defined below based on criteria in ANSI /ANS 58.8.
3 s
a
3 to = seismic event occurs causihg loss of offsite power which i
causes auto AFW initiation, and loss of control air.
ta = reactor trip due to loss of power, 2-seconds after event
. initiation.
ta = time auto AFW actuated, approximately 2 seconds after event initiation, tm = 30 minutes after automatic initiation of AFW for. action outside the control room, 30 minutes: and 2 seconds.after event initiation.
tp = time to complete safety =
function,, 15 minutes after event initiation for this event..
te = same time as tp since the operator task'is not complete until he has established the design criteria AFW flow.
ti based on 5 required manipulations, 5 minutes' for the
=
manipulations, plus 1 minute-for the operator to determine 1
the need to manually control the turbine, or 6 minutes-total for this event. The ti time is established by subtracting the-time from tp, or 9 minutes after -the loss of off-site power for this scenario.
The time line for Case 2 is shown in Figure'2.
l WALK THROUGH - CASE 1 A walk through of this event was performed in the. plant by.
simulating the conditions -with a normal shif t compliment and allowing operators using plant procedures -to respond as they-would-in the actual situation.
- Timing was started at a simulated loss of feed water, 7 seconds later the. operators were advised they-had a reactor trip on low steam generator level coincident with SF/FF mismatch.
After operators.
completed the first four steps of E-0 they. were advised 1 that ~
i SI was not actuated they would monitor critic @al When - operators indicated.
or re ired.
safety function status trees, they were advised-that a. red path on Heat Sink. existed.
Operators then proceeded to FR-H.1.
After the operators reached step 2 of.this procedure, and simulated. increasing AFW flow to 320 gpm, timing was stopped.
The total time to complete this simulation was 3 minutes and 24 seconds.
This time is considered conservative - for. the following reasons. Operators were not allowed to take action l-
' 'to restore the required AFW flow until the procedure step that addresses AFW was reached. Operators were required to read all steps of the procedure, simulate, all required
- actions, and not rush through the procedure.
In actual.
situations, the operators are allowed ' to take action after 4
I f-y t,
I
' mmediate operator
-actions.
In the completing their-i simulation, they' were forced-to wait until -the procedure'-
reader got to the applicable step. The actual procedure steps-used in the simulation are shown,in Attat.hment - 1.
Although the operators involved in :the simulation may be classed-as slightly above average, concerns over proficiency biasing of the results is minimized by requiring the operators to use procedures step by~ step.
WALK THROUGR - CASE-2 A _ walk through of this event was ; conducted 1 la a similar-manner as case 1. Timing for this event assumed that-it would take 30 seconds for the control -room operators to recognize that AFW was automatically actuated 'and page an operator to '
respond. _
The operator performing the local action was:
assumed to be in the Screen House for the worst case. Timing was started from the moment _the operator received a page in.
the Screen House.
The operator proceeded to-the Terry turbine room and was advised that he-discovered both turbines tripped on overspeed. Timing _ was stopped after the operator had simulated using the procedures to take local manual control of a Terry turbine and coordinated ' with the control '
-room to establish 320 gpm, The total time to complete this simulation was'4 minutes and 38 seconds, which included 30 seconds'to page the1 operator,.2
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minutes to proceed to the. Terry turbine room, and-2 minutes' and 8 seconds to start the Terry turbine using the= applicable procedure.
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1 Figure l-
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Loss of Feedwater Time Line (Case 1) t, te t
t, t,
ti t.
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0mSi 11 'in is ain l
7 sec 5 min 7 see m
Ioss Rx trip operator Operator Operator-of Red Path /
Action Action Action and Feed Heat sink Allowed Required safety Function.
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l Figure 2 Loss of Control Air Time Line (Case 2) t, t
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t t,
t o
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t, p
0 min
. 2 see -
9 min 15 min 30 min 2 sec j
seismio ABW Operator Operator operator
- Event Automatically Action Action and Actice ooours Lotuated Required safety Function
-Allowed ocaplete
- Per ANSI /ANS 58.8-1984 Operator Action outside the Control Room is not allowed for 30 minutes following indications that an event has occurred.
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-l.i.I ATTACHMENT l
PROCEDURES.USED DURING WALK-THROUGHS l
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CASE.i SthtAnie4 Pttoc.eM.Er REV. ISSUB/DA'!E
.I NUMBER Trn.B I
E.0 REACTOR TRIP OR SAFETY INJECTION Rev. 7 OYb W
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[dIEP ACTION /EXFBCTED RESPONSB RESPONSB NOT OBTAINED vs N
% NOTEt o
STEPS 1 THROUGH 13 ARE IMMEDIATE ACTION STEPS.
T o FOLDOUTPAGE SHOULD BE OPEN.
.,, l' EPIP 1.51, EMERGENCY ASSESSMENT, SHOULD DE IMPLEMENTED.'
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'. VERIFY REACTOR TRIP t o
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- a. Push reactor trip button s/
- b. Check the following: -
b, QQ TQ FR-S.I.
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/j o Rod bottomlights-LIT RESPONSETO NUCLEAR e
- o Reactor trip breaker (s)- OPEN POWER GENERATION /ATWS, o Neutron flux-DECREASING Step 1.
i 2
VERIFY TURBINETPJP:
a.iPush turbine trip button I: a
- b. Check the following:
- b. Close allfourmain steamilne
- 1. Both turbine trip valves - CLOSED trip valves.
.QR
.GQIQ Step 3.
d
- 2. Allfourturbine govemorvalves-CLOSED p!
- c. Check turbine speed -
- c. Close all four main stearallne sI LESS THAN 2300 RPM trip valves.-
3
' VERIIW POWERTO ACEMERGENCY BUSES:
/
- a. Check AC emergency buses energized -
- a. Attempt to restom power to at
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AT LEAST ONE TRAIN least one train.
-IE NOIreston:d, Buses 8,4 and 5 THT!N GQ TQ ECA 0.0,
.QR STATION BLACKOUT, Step 1.
Buses 9,6,7 and 11
- b. Check AC emergency buses energized -
- b. Attempt to restore power to BOTH TRAINS deenergized AC emergm:y haet.
MIDcontinue with Step 4.
Buses 8,4 and 5 AND Buses 9,6,7.and 11 4
CHECKIF SI ACTUATED:
CHECKIF SIIS REQUIRED.
o Check SIWL Relays -
'DJEM MANUALLY ACTUATE.
CNE OR BOTHTIUPPED IE SI NOT REOUIRED, THEN QQIQ ES-0.
g REAUrUR TRIP SPONSE, STEP 1.
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NUMBER TITLB REV. ISSUB/DATB FR.II 1 RESPONSE TO LOSS OF SECONDARY Rey, 7 IIEAT SINK O.',
l ACTION /BXPECTED RESPONSE
. RESPONSB NOT OBTAINED STEP 7
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.b #g CAUTIONt o
IF TOTAL FEEDWATER IS LESS THAN 320 GPM DUE TO OPERATOR ACTION, THIS PROCEDURE SHOULD NOT BE C'
@ ch PERFORMED.
l IF EITHER OF THE FOLLOWING CONDITIONS EXIST, RCP.s o
SHOULD BE TRIPPED AND STEPS 10 THROUGH 15 SHOULD DEIMMEDIATELYINITIATED FOR RCS BLEED AND FEED:
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o CORE EXIT THERMOCOUPLES ARE APPROACHING 575
- F AND. RCS PRESSURE IS APPROACHING 2350 PS10.
OR o
WIDE RANGE LEVEL IN ALL FOUR STEAM GENERATORS M
15 LESS THAN 21 %.
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FEED FLOW SHOULD NOT BE ESTABLISHED TO ANY FAULTED SG IF A NONFAULTED SG IS AVAHARLE.
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1 CHECKIF SECONDARY HEAT SINK IS REQUIRED:
- a. Check RCS aressure-
- n. QQIQ E 1, GREATER'TIAN ANY LOSS OF REACTOR OR s
NON FAULTED SG PRESSURE SECONDARY COOLANT, Step 1.
- h. Check RCS temperature -
- b. Try to place RHR system in service OREATER'lilAN 300 F IN ACCORDANCE WITII
(
(277 FFOR ADVERSE NOP 2.9-1, CONTAINMENT)
RHR SYSTEM PLACING SYSTEM IN SERVICE, while continuing in this procedure.
IE adec unte core cooling established with RAR System in service, THEN RETURNIQ procedure and step in effect.
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NUMBER TRLE REV. ISSUB/DA'I11
- FR II.1 RESPONSE TO LOSS OF SECONDARY Rey, 7 JIEAT SINK
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STEP ACTION /BCEC"MID RBSPONSB RBSPONSB NOT ObTAINBD -
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NOTE JF BOTH TURBINE DRIVEN AFW PUMPS ARE NOT OPERABLE, THEN THE ELECTRIC AFW PUMP CAN BE LOCALLY AUONED, IN ACCORDANCE WITH NOP 2.18 3, ELECTRIC AUX 1UARY FEED PUMP OPERATION, TO PROVIDE AFW FLOW.
2 TRY TO ESTABLISH AFW FLOW -
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TO ATLEASTONE SO:
- a. Checkcontrolroomindications for cause of AFW failure:
.J o-CheckDWSTlevel o
Check AFW valve alignment gg uanarA
- fWM j m o uc e W 'f # g ro M E
- b. Try to restore AFW flow W *Tb dtD g m.
3
- c. Check total AFW flow-
- c. Checkthe A turbinedriven APW GREATER THAN 320 GPM pump operating,
.DiFM perform the following:
C J6 L9 sec. ode
- 1. Close FW MOV-160.
~Mg.ng.
- 2. Open FW MOV-35.
m3.s rror um.g cyit S 171
- IEFW MOV-35 falls to own,
cA b#: 5* u 9 TiiEN perform the follow,ng:.
cg Dog, ' n l. 2. l* 90 '
- a. Sto the A AFWpump, b..O n FW-MOV 35.
- c. Restart the A AFW pump. *2 j
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- 3. Check SO levels increasing.
OR Sd$vefs'NQIincrea:In,
THEN dispatch operator to locally -
restore AI4W flow.-
OQIQ Step 3.
- d. RETURNIQprocedure and step in effect C
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AOP 3.2 51-s /;
Rev. 2 -
f ConnecdentYankee Abnormal Operating Procedure Local Manual Operation of the Auxillary Feettwater System 1.0 DTSCUSSION b
1.1 The auxillary feed pumps are normally operated from the Control Room. %ey may'
- i be operated locally, if necessary. Delr operation ensures that water levels are Should a loss of control air occur, the main feed regulating valves cl maintained in the Steam Generators providing a heat Ink for the Primar the feedwater i
bypass valves open, and the steam supplycontrol valves for % auxiliary feed pumps open.
2.0 SYMPTOMS 2.1 Steam Generatorlow level alarms.
L 2.2 Control airlow pressure alarms.
l 2.3 dontrol Room fire, loss of control air, or controller malfunction.
'3.0 AUTOMATIC ACTIONS 3.1.
N/A.
i 4.0 OPERATOR ACTTONS, NOTE-Establish communications between the dirpatched operator l
and the operator monitoring Steam Generator levels, J
4.1 Start the Auxiliary Feed Pumps using Local Operation' M
.o 4.1.1 DISPATCH o rator to Terry Turbine Room with instructions to manually -
4' operate auxill feed pump and the feedwater path to use, p'b CLOSE the auxillary feed p steam supply control valve by 711RNING o
the handwheel in the cloc s direction untilit stops, n.
For P 321 A, CLOSE steam supply control valve (MS PICV 1206A),
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For P-321B, CLOSE steam supply control valve (MS PICV 1206B).
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TF necessary, RESET the auxillary feed pump overspeed trip by j'
',C ENOAOINO the trip rod.
regulator.CLOSE the air supply isolation valve to the ]
a.
CLOSE CA V 258 (located on wall by steam pressure Indication) for P 321 A.
~
b.
CLOSE CA-V 259 (located by AOV) for P 32-1B, :
h OPEN the air regulator draln to DEPRESSURTZE the steam' supply co valve air supply.
Q ShART the auxillary fe'ed pump (s) by slowly TURNTNO the steam supply control valve handwheelin the colmrer clockwI.te direction:
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req 8ToHU "*'
For P 321A, OPEN steam supply control valve (MS PTCV 1206A) to a.
pm rrhr y establish the pump's discharge pressure at approximately 1100 psig..
CwwW'. C8T**' ** * *
- b. - For P 321B, OPEN steam supply control valve (MS PTCV 1206B)'to f
p, cpau. W w establish the pump's discharge pressure at approximately 1100 psig.
4.1.7 VERTFY discharge to the feedwater station (FW V 157) OPEN.
1
, TarM *r A' 4 tim M 8*' ' 4.1.8 VERIFY FW MOV 160 is OPEN and FW MOV 35 is CLOSED.
4.2 Feed Individual Steam Generators 4.2.1 CHECK OPEN intet isolation valves to the feed bypass control valves (FW V-1571,2,3, & 4).
4.2.2 To ensure that the feed bypass control valves are OPEN and remain open, PERFORM the following:
a.
UNLOCK and REMOVE the locking device from the three way valve located on top of the feed bypass control valve, b.
ROTATE the vent valve on top of the feed bypass control valve to bleed air from the diaphragm.
4.2.3 Manually REGULATE flow to the Steam Generator by CLOSTNO then THRO'ITLING OPEN the feed bypass control valve outlet isolation valves two (2) hand turns.
4.2.4 ESTABLISH communications with the operator monitoring Steam Generator levels and REOULATE feedwater flow to maintain proper Steam Generator levels.
4.2.5 When the level in the DWST appmaches 54,000 gallons, COMMENCE filling the tank by selecting one of the following methods:
n, To fill the DWST fmm the PWST, PROCEED to Section 4.4.
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1, Docket No. 50-211 A08995
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' Attachment No. 2 6
Haddam Neck Plant Additional Information Design Basis /Best Estimate Analysis-Assumptions l
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6 September 1990
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A08995/Page 1 l
Haddam Neck Plant Design Basis vs. Best Estimate Analysis Assumptions Desian Basis Best~ Estimate Power Level 102%
100%
RCS Temperature Minimum Nominal Condenser Unavailsble Available-AFW Flow 255 gpm 0 4 minutes
'165 gpm 0 1.5 minutes (l' pump)-
265 gpm.0 15 minutes
.(1 pump) e tr M imu Temperature AFW Temperature
~125'F 80*F
- Note:
This list. represents -most of the significant differences in the assumptions of the two scenarios.
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September 1990