ML20059H553
| ML20059H553 | |
| Person / Time | |
|---|---|
| Site: | 05200002 |
| Issue date: | 11/03/1993 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC. |
| To: | |
| Shared Package | |
| ML20059H479 | List: |
| References | |
| PROC-931103-01, NUDOCS 9311100122 | |
| Download: ML20059H553 (37) | |
Text
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SYSTEM 80 + " TITLE REACTOR TRIP RECOVERY c EMERGENCY OPERATIONS GUIDELINES page 5 of n Revision "" I
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REC 0VERY GUIDELINE i
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9311100122 931103 P; y i PDR ADOCK 05200002 PDR Li A i
SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS Page 2 of Revision "
GUIDELINES PURPOSE l 1
This guideline provides the operator actions which must be accomplished :
subsequent to a relatively uncomplicated reactor trip. The actions in this i
guideline are necessary to ensure that the plant is placed in a stable, safe condition. The goal of the guideline is to safely establish the plant in a mode 3 condition (hot standby) while minimizing any radiological releases to the environment. If necessary, the RCS may be cooled and depressurized.
i ENTRY CONDITIONS
- 1. The Standard Post Trip Actions have been performed and
- 2. Plant conditions indicate that an uncomplicated reactor trip has occurred.
I EXIT CONDITIONS i
- 1. The diagnosis of an uncomplicated reactor trip is not confirmed .
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- 2. Any of the Reactor Trip Safety Function Status Check acceptance criteria are not satisfied ,
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- 3. The Reactor Trip Recovery E0G has accomplished its purpose by satisfying ALL of the following: ,
- a. All Safety Function Status Check acceptance criteria are being satisfied.
- b. RCS conditions are being controlled and maintained in a mode 3 or 4 condition (hot standby or hot shutdown).
- c. An appropriate procedure to implement has been provided and
' administratively approved.
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SYSTEM 80+" TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS Page ' of Revision ^^" ,
GUIDELINES ,
INSTRUCTIONS CONTINGENCY ACTIONS ,
i
- 1. Ensure Standard Post Trip 1.
Actions performed. ;;
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- 2. Confirm diagnosis of 2. Rediaonose event and exit to uncomplicated Reactor Trip by appropriate Optimal Recovery verifying Safety Function Status Guideline or to Functional -
Check acceptance criteria are Recovery Guideline.
satisfied. l t
- 3. Verify pressurizer level is: 3. Manually operate PLCS or
- a. [2-78%] charging and letdown to restore {
and and maintain pressurizer level
- b. trending to [33-52%] [33-52%]. t
- 4. Verify pressurizer pressure is: 4. Manually operate PPCS or 'i pressurizer heaters and spray to j
- a. [2000 to 2300 psia] control RCS pressure: i
- a. [2225 to 2300 psia]
and
- b. trending to [2225 to 2300 and ;
psia] b. within the Post Accident P-T }'
and limits of Figure 4-1.
r
- c. within the Post Accident P-T >
limits of Figure 4-1. {
SYSTEM 80 +" TITLE REACTOR. TRIP RECOVERY l EMERGENCY OPERATIONS !
GUIDELINES Page ' of 37 Revision '""
INSTRUCTIONS CONTINGENCY ACTIONS
- 5. Verify steam bypass control 5. If condenser vacuum is lost, system is controlling RCS T,,, steam bypass control system is ,
[551-562*F). unavailable, or the MSIVs are closed, Then use the atmospheric dump valves to control RCS T,.,
[551-562*F].
- 6. Ensure at least one steam 6.
generator has level being maintained or restored in the normal band using main, startup or emergency feedwater.
- 7. Evaluate the need for a plant 7.
cooldown based on:
- a. plant status
- b. auxiliary systems availability
- c. condensate inventory.
- 8. If a plant cooldown is 8.a. Maintain the plant in a necessary, Then exit this stabilized condition, guideline and implement the and appropriate plant cooldown b. Exit to appropriate procedure, procedure as directed by
[ Plant Technical Support Center).
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SYSTEM 80 + C TITLE REACTOR TRIP REC 0VERV EMERGENCY OPERATIONS Page 5 of 37 Revision ^^"
GUIDELINES When the steps of the Reactor Trip Recovery Guideline are complete, the plant -
should be in a condition where all of the SFSC acceptance criteria are satisfied, and the entry conditions of an appropriate procedure are satisfied. -
In most cases, the plant will be maintained in hot standby or directed to be cooled down to mode 4 or 5.
END ;
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I SYSTEM 80+" TITLE REACTOR TRIP REC 0VERY !
EMERGENCY OPERATIONS GUIDELINES Page ' of " Revision ""
SUPPLEMENTARY INFORMATION ,
This section contains items which should be considered when implementing E0Gs -
and preparing plant specific E0Ps. The items should be implemented as +
precautions, cautions, notes, or in the E0P training program, t
- 1. Pressurizer level should be closely monitored since it normally decreases to, or near, the pressurizer heater cutoff level following a reactor trip. r
- 2. All available indications should be used to aid in evaluating plant conditions since the accident may cause irregularities in a particular instrument reading. Instrumentation readings must be corroborated when one or more confirmatory indications are available, (e.g., during rapid ,
depressurization the indicated level in the pressurizer may be too high).
- 3. A plant cooldown and entry into shutdown cooling (if necessary) should be i conducted prior to depleting the condensate storage. !
- 4. During all phases of the cooldown, RCS temperature and pressure should be monitored to avoid exceeding a cooldown rate greater than Technical Specification limitations. ;
- 5. Do not place systems in " manual" unless misoperation in " automatic" is )
I
, apparent. Systems placed in " manual" must be checked frequently to ensure proper operation.
- 6. If the initial cooldown rate exceeds Technical Specification Limits, then there may be a potential for pressurized thermal shock (PTS) of the reactor vessel. Post accident pressure / temperature should be maintained within the limits of Figure 4-1.
SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS 7 GUIDELINES Page of 37 Revision "
1 t
C Figure 4-1 Typical Post Accident Pressure-Temperature Limits ,
(TO BE DEVELOPED DURING DETAILED ENGINEERING) l t
SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS 8 GUIDELINES Page of 57 Revision "
SAFETY FUNCTION STATUS CHECK ;
SAFETY FUNCTION ACCEPTANCE CRITERIA
- 1. Reactivity Control 1.a. Reactor power decreasing and
- b. Negative Startup Rate ;
and
- 2. Maintenance of Vital Auxiliaries 2.a. Safety Load Division I ener-(AC and DC Power) gized via Permanent Non- 1 Safety Bus X E
Safety Load Division II energized via Permanent Non-Safety Bus Y ,
and
- b. Non-safety load [13.8 KV)
Bus X energized '
E Non-safety load [13.8 KV]
Bus Y energized and ,
- c. Non-safety load [4.16 KV]
Bus X energized i E
Non-safety load [4.16 KV]
Bus Y energized and RT 8 ABB CE SYSTEM 80+"
SYSTEM 80+" TITLE REACTOR TRIP REC 0VERY EMERGENCY OPERATIONS GUIDELINES Page ' of 3' Revision ""
SAFETY FUNCTION ACCEPTANCE CRITERIA 1
- 2. (Continued) d. Permanent Non-safety load
[4.16 KV] Bus X energized -
or Permanent Ibn-safety load
[4.16 KV] Bus Y energized and ;
Safety Bus A energized r
E i
[125V] DC and [120V] AC Safety Bus B energized E !
[125V) DC and [120V] AC Safety Bus C energized E
[125V] DC and [120V] AC Safety Bus D energized
- 3. RCS Inventory Control 3.a. Pressurizer level is [2%to 78%]
i and
- b. Charging and letdown are restoring pressurizer level ;
to [33%to 52%)
and
- c. The RCS is subcooled and
- d. No reactor vessel voiding as ;
indicated by the RVLMS. j i
SYSTEM 80+" TITLE REACTOR TRIP RECCVERY-EMERGENCY OPERATIONS GUIDELINES.
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SAFETY FUNCTION ACCEPTANCE CRITERIA
- 4. RCS Pressure Control 4.a. Pressurizer pressure is:-
2.1 WJ ^
i) [2000-2300 psit
and ii) trending to [2225 to 2300 psia]
and
- b. Pressurizer heaters and spray are controlling pressure within P-T limits ,
of Figure 4-1.
- 5. Core Heat Removal 5.a. Tu - Tc is less than [3*F]
and
- b. The RCS is subcooled.
- 6. RCS Heat Removal 6.a. At least one SG has level: j i) within normal level band with feedwater available to maintain level ,
or ;
ii) being restored by McdS s U4 ,'
or em feedwater flow l
grect:r-ther. [50 sp- ,
5 r Leani scuc.eter'r ,
and increasing level and
- b. RCS T,,,is (551-562*F].
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GUIDELINES SAFETY FUNCTION -ACCEPTANCE CRITERIA
- 7. Containment Isolation 7.a. Containment pressure less than [2.0 psig] :
and i
- b. No containment area radiation monitors alarming and ,
- c. No steam plant radiation '
monitors alarming i and
- d. No nuclear annex radiation .
monitors alarming and
- e. No reactor building radiation monitors alarming j
- 8. Containment Temperature & 8.a. Containment temperature.less Pressure Control than [110*F]
and ,
- b. Containment pressure less than [2.0 psig].
t
- 9. Containment Combustible Gas 9.a. Containment temperature less Control than [110*F]
and
- b. Containment pressure less than [2.0 psig]. i E
SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS Page '2 of " Revision ""
GUIDELINES BASES-The bases-section of the Reactor Trip (RT) Recovery Guideline describes the RT transient in relatior, to the actions which the operator takes during the l recovery from a relatively uncomplicated RT. The purpose of the bases section is to provide the operators with information which will enable them to understand the reasons for, and the consequences of, the actions they take during a RT. !
Characterization of a Reactor Trio A reactor trip is a shutdown of the reactor accomplished by the rapid ;
insertion of the control element assemblies (CEAc). It is automatically-initiated by the reactor protective system when certain continuously monitored parameters exceed predetermined setpoints, or it can be initiated manually by the operator if plant conditions warrant. A malfunction in the reactor protective system may also cause a reactor trip signal. ,
A reactor trip may be the result of automatic action initiated by the reactor protective system in response to any of the following .;. 4ical parameters: ,
- a. High reactor power.
- b. Low pressurizer pressure.
- c. Low reactor coolant flow. ,
- d. Low steam generator level.
- e. Low steam generator pressure.
- f. High pressurizer pressure.
- g. High steam generator level.
- h. High containment pressure.
- i. Turbine trip. :
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- j. DNBR trip.
- k. LPD trip. ;
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SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY !
EMERGENCY OPERATIONS GUIDELINES Page " of " Rev..ision "" i Safety Function Affecte_d_
A reactor trip results in a decrease of primary system heat generation to !
decay heat levels. It is a safety action performed for reactivity control and ,
does not directly challenge the maintenance of any safety function required to !
place the plant in a safe, stable condition. However, all safety functions i should be monitored to assure public safety or to detect failures which may lead to unsafe conditions.
Trendina of Key Parameters j Reactor Power (Figure 4-2) -
As a result of the reactor trip, the control element assemblies (CEAs) will be ,
rapidly inserted. Steam flow to the turbine generator will be terminated, the ;
turbine generator output breakers will open and the feedwater flow will l automatically ramp down-te eppr= int:1.7 % now. A rapid det.rease in reactor j power and a negative startup rate will be observed. This rapid decrease is l followed by a decrease in indicated power (approximately -1/3 decade per l minute) until the subcritical multiplication level is reached. Indicated ;
a power will stabilize at the subcritical multiplication "el and decrease slowly over a period of hours. i RCS Temperature (Figure 4-3) ]
Initially, feedwater temperature decreases sharply due to the loss of steam heating to the feedwater hgaters (450*F to about 200*F feedw:.ter temperature) i 1
or due to actuation of =%a t iary e emem(may feed be as low as 40*F). Heat from the RCS is absorbed by the cooler feedwater supplied to the steam generators. At power, there is a large differential between RCS Tave and average steam i generator temperature. Following the trip of the reactor and the turbine, the heat transfer rate from the RCS to the steam generator decreases to decay heat l removal levels and the RCS to steam generator 4T decreases to a few degrees )
with RCPs running.
i SYSTEM 80 +" TITLE REACTOR TRIP REC 0VERV EMERGENCY OPERATIONS Page " of " Revision ""
GUIDELINES As a new equilibrium is achieved, the combined effect of the cooler feedwater and the steam generator heating up to an average temperature closer to RCS temperature results in a net heat extraction from the RCS. Loop differentials ;
between hot and cold leg temperatures will drop to less than ten degrees with j RCPs running and RCS average temperature will decrease to [551-562*F]
- controlled by the steam bypass control system or the atmospheric dump valves.
Reactor Vessel level j t
For an uncomplicated reactor trip, it is expected that the reactor vessel will l remain full. The subcooled margin in the RCS loops is typically [50*F] or higher, and RVUH subcooling margin can be significantly lower than that for the RCS loops but still high enough to prevent voids froin forming. At steady l state conditions, the upper head region is about l'F cooler than the core exit 4 temperature and, therefore, the subccoled margin of the RVUH is essentially i equal to that of the hot leg. Under transient conditions, with RCPs running, j there is a time lag between the change in the core exit temperature and the .
- change in RVUH temperature to approximately the same temperature. Under RCS !
cooling transients up to [75'F/ hour], the time lag is small enough so that the -
subcooling margin in the RVUH will not allow voids to form. ,
Pressurizer Pressure and level (Figures 4-4, 4-5) l Pressurizer pressure and level will initially decrease due to the lowering of RCS temperature. However, this effect will usually be tempered by operation of pressurizer heaters and the charging pump which restore level to the i
programmed hot zero power band.
1 Steam Generator Pressure (Figure 4-6)
Steam generator pressure will usually increase. Since heat is being removed t
from the RCS but not from the steam generator (except for the cooling from the -
t feed), the steam generator heats up to decrease RCS to steam generator RT 14 ABB CE SYSTEM 80+"
4 SYSTEM 80 +" TITLE REACTOR TRIP REC 0VERY !
EMERGENCY OPERATIONS Page " of '7 Revision " ;
GUIDELINES differential temperature. Steam generator pressure increases as temperature increases. As steam generator pressure increases, the turbine bypass valves will usually open or the atmospheric dump valves will be opened to control steam generator pressure at hot standby pressure (which is above normal 100% ;
power steam generator pressure), f Steam Generator level (Figure 4-7)
After a reactor trip the steam generator level decreases rapidly. This is i
explained as follows. Steam generator level is inferred from the steam generator downcomer level. During normal 100% power operation, the steam generator has a recirculation ratio of approximately 4 to I (ratio of water ;
returning to the downcomer from the dryers and separators to feedwater i entering the downcomer). This accounts for a major portion of the water level j entering the downtomer. When steam flow is stopped by the turbine trip, '
recirculation stops. The reduced flowrate into the downcomer results in reduced head losses through the downcomer and up the riser section. The downcomer water level, and thus the steam generator indicated level, both drop. This drop in level will occur even before the feedwater system i
automatically readjusts. Another contributing factor to the observed decrease in post trip S/G water level is the increased steam generator pressure. This increase in pressure causes an increase in the saturation temperature, thus ;
causing the voids in the S/G to collapse.
Plant operators should be cautioned not to overreact to this lowered level in the steam generators. Excessive feeding of the steam generator with cooler I
feed to recover level results in RCS temperatures being driven down below the desired no load value. This could cause pressurizer level to fall to a point ,
where the pressurizer is drained. RCS pressure will then drop until the ;
safety injection system is actuated. This complicates the recovery from a simple reactor trip considerably. r i
- - - _ _ - - - - - _ _ _ _ _ 1
1 SYSTEM 80+" TITLE REACTOR TRIP RECOVERY- l 1
EMERGENCY OPERATIONS Page " of Revision """
GUIDELINES i
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Figure 4-2 REPRESENTATIVE REACTOR TRIP REACTOR POWER (TO BE DEVELOPED DURING DETAILED ENGINEERING) r t
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SYSTEM 80+" TITLE -REACTOR TRIP RECOVERY EMERGENCY OPERATIONS GUIDELINES P a g e of Revision ""
i h
Figure 4-3 ;
REPRESENTATIVE REACTOR TRIP !
RCS WIDE RANGE TEMPERATURES i (TO Br. DEVELOPED DURING DETAILED ENGINEERING) ,
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SYSTEM 80 + " TITLE REACTOR TRIP REC 0VERY EMERGENCY OPERATIONS GUIDELINES Page
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b s
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Figure 4-4 REPRESENTATIVE REACTOR TRIP PZR WIDE RANGE PRESSURE f i
(TO BE DEVELOPED DURING DETAILED ENGINEERING) ,
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SYSTEM 80 + = TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS Revision ""
Page " of '7 GUIDELINES Figure 4-5 REPRESENTATIVE REACTOR TRIP PZR LEVEL (TO BE DEVELOPED DURING DETAILED ENGINEERING)
SYSTEM 80 + = TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS 2" Page of Revision "
GUIDELINES F
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Figure 4-6 REPRESENTATIVE REACTOR TRIP STEAM GENERATOR PRESSURE (T0 BE DEVELOPED DURING DETAILED ENGINEERING) l RT 20 ABB CE SYSTEM 80+"
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SYSTEM 80 +" TITLE REACTOR TRIP. RECOVERY i
EMERGENCY OPERATIONS GUIDELINES P a g e of Revision ""
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Figure 4-7 REPRESENTATIVE REACTOR TRIP STEAM GENERATOR WIDE RANGE LEVEL (TO BE DEVELOPED DURING DETAILED ENGINEERING) ,
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SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY l EMERGENCY OPERATIONS Page 22 of 37 Rev..ision ""
GUIDELINES Guideline Strategy ,
Figure 4-8 provides a summary of the Reactor Trip (RT) Recovery Guideline's strategy. Prior to implementing the actions provided in the RT Recovery l Guideline, the operator would have performed the-Standard Post Trip Actions .
and concluded that an uncomplicated reactor trip had occurred. In the RT Recovery Guideline the operator begins using the Safety Function Status Check to confirm that the plant is recovering and the correct guideline has been implemented. RT Recovery actions provide instructions on regaining and i
maintaining RCS inventory control, RCS pressure control, and RCS heat removal.
l l A more detailed RT Recovery strategy chart is provided. It lists the guideline steps which correspond to each strategy objective. Refer to Figure 4-9.
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SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS GUIDELINES Page " of " Revision ^^"
FIGURE 4-8 REACTOR TRIP g-_
g STAS 0ARO PSIT1 RIP !
l Pt ResEO _,
F ~ UNCO l CAT E '!
R EACTO R T RIP F
l DIA850310 J
$PTAs AND DIAGNOSTit ACT50kt ,
REACTOR TRIP RECOVERY ORG qr
[
vf3 /EI PN 40 9P APPROPGLATE OR$
OR FRG 46 1P YE5 0 48 STEC SATt1FitD 1
1P CCNTROL RCS INvis-TORY Amo PRESSURE 1r CONTROL RCS HEAT REMOVAL t
l W
vts C00LC he l
iP 1r MAINTAIN PLAET IN '
COOLD0wsTO3CS ET A8LE Countil0E INTRY CCtDITtom8 t i
SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS Revision ""
Page 2' of '7 GUIDELINES Bases for Operator Actions The operator actions are directed at bringing the plant to a safe, stable condition following an uncomplicated reactor trip and ensures that a proper heat sink for the reactor is being maintained.
- 1. The operator is directed to ensure that the Standard Post Trip Actions have been performed. This action ensures that all safety functions have been monitored, and appropriate contingency actions performed, prior to implementing the Reactor Trip Recovery EPG.
- 2. The operator is required to continually verify that Safety Function Status Check acceptance criteria are satisfied by comparing control board parameters to the acceptance criteria in the Safety Function Status Check. This ensures that the safety functions are satisfied and the core is being adequately cooled. If the Safety Function Status Check acceptance criteria are satisfied adequately mitigating the effects of the RT. Thus, the implementation of the remaining actions of this guideline is continued. If the diagnosis of an uncomplicated reactor trip is found to be in error (i.e., any of the Safety function Status Check acceptance criteria are not satisfied), the procedure is not adequately mitigating the event. If another event is diagnosed, the operator exits the RT guideline and implements the appropriate Optimal Recovery Guideline (ORG). If a diagnosis of one event cannot be made, the Functional Recovery Guideline (FRG) is implemented. The FRG is safety function based and will ensure all safety functions are addressed regardless of what event (s) is occurring.
- 3. Following a relatively uncomplicated reactor trip, pressurizer level should not decrease below [2%] or increase above [78%]. The value of
[78%], was chosen as an upper limit for pressurizer level to account for some process fluid uncertainties in order to avoid solid plant operation. The process uncertainties include maintaining an operable RT 24 ABB CE SYSTEM 80+"
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SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY j 1
EMERGENCY OPERATIONS 3 of y Page Rev..ision ,,
GUIDELINES ;
steam bubble following a 20 second inadvertent initiation of auxiliary spray. The value of [2%] was chosen as the lower limit in order to l avoid draining the pressurizer. If the pressurizer level control system functions properly, level in the pressurizer should be trending to [33- l 52%]. This will ensure that the PLCS is working properly to control level. If automatic pressurizer level control system operation is not maintaining / restoring level, the operator is instructed to take manual control of charging and letdown to control pressurizer level to [33- ,
52%).
- 4. Following a relatively uncomplicated reactor trip, automatic control of i pressurizer heaters and spray should be sufficient to maintain pressurizerpressure[b-2370 psia]. The availability of pressurizer ,
heaters will be dependent upon the ability to restore pressurizer ziLo level !
to above the heater level of [14.3%). The lower value of [te98J psia] j corresponds to the RCS low pressure alarm setpoint. The higher value of
[2370 psia] is the RCS high pressure alarm setpoint. If the pressurizer pressure control system functions properly, pressurizer pressure should j It be trending to [200*F] psia. Satisfying the Post Accident P-T limits of ,
h Figure 4-1 will ensure that brittle fracture limits are not exceeded, f RCP NPSH and RCS subcooling requirements are satisfied, and RCS cooldown j i
rate 100*F/ Hour or upper subcooling limit [200*F] are not exceeded. If automatic pressurizer pressure control system operation is not maintaining / restoring pressure, the operator is instructed to take manual control of pressurizer heaters and spray to control pressurizer f pressure to [2225-2300 psia]. [
- 5. RCS T,,, should be controlled at [551-562*F] by the steam bypass control l system. If condenser vacuum is lost, the steam bypass control system is .
not available, or the MSIVs have closed, the atmospheric dump valves i must be used to control RCS T,,,in a [551-562*F] band. This step provides a verification that the steam generator (s) are adequately RT 25 ABB CE SYSTEM 80+*
1 SYSTEM 80 + " TITLE REACTOR TRIP RECOVERY >
i EMERGENCY OPERATIONS Page 2' of " Revision ""
GUIDELINES removing decay heat, that control systems are functioning properly, or that manual actions are taken as appropriate. ;
- 6. Following a relatively uncomplicated reactor trip, steam generator levels should automatically be restored and maintained in the normal i level band. The operator will ensure that automatic or manual control of main, startup, or emergency feedwater is capable of maintaining at ;
least one steam generator's level in the normal band. Adequate RCS heat ;
removal will be maintained if at least one steam generator is available i for removing heat (capable of feed and steam flow). The operator must use caution when manually feeding steam generators to avoid an excessive !
RCS cooldown rate with subsequent pressurizer level and pressure transient or overfilling steam generators. Steam generator levels-should be increased at a rate consistent with decay heat levels and any desired cooldown rate. A flot ate of [250 gpm per steam generator] is .
sufficient feed flow to remove decay heat.
- 7. At this point in the recovery, the operator should determine whether a plant cooldown is necessary. If the continued availability of any systems required for maintenance of HOT STANDBY is in doubt, a cooldown may be appropriate. For example, if the available condensate inventory j
. is marginally adequate, a cooldown should be performed in order to avoid l running out of condensate before the shutdown cooling system can be placed into operation. Similarly, consideration should be given to the ,
availability of compressed air and cooling water systems as well as the i continued availability of electrical power. A cooldown may also be required in order to provide the plant conditions necessary to perform system or component repairs.
- 8. If a plant cooldown is necessary and the RT Recovery Guideline exit conditions are satisfied, this guideline should be exited and the plant cooldown procedure implemented. If it is decided that a cooldown is not necessary, the plant should be maintained in a stable condition until RT 25 ABB CE SYSTEM 80+"
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i SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY ;
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GUIDELINES i
the operators and the support staff [ Plant Technical Support Center] l determine which procedure is appropriate to implement.
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Safety Function Status Checks ]
The Safety Function Status Check (SFSC) is used to continually verify the status of safety functions. The safety function acceptance criteria are selected from best estimate analysis to reflect the range for each parameter which would be expected following a relatively uncomplicated reactor trip. If ,
all SFSC acceptance criteria are being satisfied, the adequacy of this guideline for mitigating the event in progress is confirmed and the health and
+
safety of the public is ensured.
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SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS Page " of 37 Revision "
GUIDELINES SAFETY FUNCTION ACCEPTANCE CRITERIA BASES
- 1. Reactivity Control a. Reactor Power Decreasing For all emergency events, and the reactor must be shut-
- b. Negative Startup Rate down. Decreasing reactor and power is one positive
- c. Maximum of I CEA NOT indication that reactivity fully inserted or the control has been estab-RCS is borated per Tech. lished. A negative startup Specs. rate can be used in the short-term post trip to verify that reactivity con-trol is established. The Technical Specification requirement is that not more than I rod be stuck out. If more than 1 rod is stuck out, the RCS must be borated to compensate for the negative reactivity not inserted into the core.
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SYSTEM 80+" TITLE REACTOR TRIP REC 0VERY -l EMERGENCY OPERATIONS Page 3, of 37 Rev..ision -
GUIDELINES
)
i ACCEPTANCE CRITERIA BASES ~ i SAFETY FUNCTION
- 2. Maintenance of vital a. Safety Load Division I One safety division of AC l Auxiliaries (AC and DC energized via Permanent power is required to power power) Non-Safety Bus X equipment necessary to or maintain control of all Safety Load Division II other safety functions.
energized via Permanent Non-safety [13.8KV]and Non-Safety Bus Y [4.16 KVJ AC divisions must .
and be available to power at l
- b. Non-safety load [13.8 least one RCP and any other KV] Bus X energized non-vital equipment E typically used in'the Non-safety load [13.8 recovery from a relatively KV] Bus Y energized uncomplicated reactor trip.
and One safety division of DC
- c. Non-safety load [4.16 is required as a minimum to KV] Bus X energized provide monitoring and or limited control of the Non-safety load [4.16 other safety functions. l KV] Bus Y energized
~
and
- d. Permanent Non-safety load [4.16 KV] Bus X energized ,
E Permanent Non-safety ;
load [4.16 KV] Bus Y energized }
I RT 30 ABB CE SYSTEM 80+* l i
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- SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS GUIDELINES P a g e of '7 Revision ""
SAFETY FUNCTION ACCEPTANCE CRITERIA BASES j
- 2. Maintenance of vital and Auxiliaries (AC and DC e. [125V] DC and [120V] AC power) Safety Bus A energized (Continued) of
[125V] DC and [120V] AC Safety Bus B energized E
Safety Bus C energized E
[125V] DC and [120V] AC Safety Bus D energized
- 3. RCS Inventory Control a. Pressurizer level is [2 The value of [78%] was to 78%) chosen as an upper limit to and ensure that the pressurizer
- b. Charging and-letdown are has an operable steam restoring pressurizer bubble. A value of [2%],
level to [33 to 52%) was chosen as the lower and limit to ensure that the
- c. The RCS is subcooled pressurizer is not drained.
based on Tg RTD ,
temperature Following a relatively and uncomplicated reactor trip,
- d. No reactor vessel automatic or manual control voiding as indicated by of charging and letdown the RVLMS. should be sufficient to '
maintain RCS inventory ,
control within [33-52%) ;
1
SYSTEM 80 +" TITLE REACTOR TRIP REC 0VERY EMERGENCY OPERATIONS Page 32 of 37 Revision ""
GUIDELINES SAFETY FUNCTION ACCEPTANCE CRITERIA BASES i
- 3. RCS Inventory Control A subcooling margin co- '
(Continued) existing with adequate pressurizer level indicates RCS inventory control via a ,
saturated bubble in the pressurizer. Ts RTDs are to be used during forced circulation flow conditions.
For an uncomplicated re-actor trip, reactor vessel voiding should not result.
7J k o
- 4. RCS Pressure Control a. Pressuggerpressureis: The lower value of [2000-i) [2000-2370 psia] psia] corresponds to the and RCS low pressure alarm 6 ii) trending to [2225- setpoint. The higher value 2300 psia] of [2370 psia) is the high and pressure alarm setpoint.
- b. Pressurizer heaters and Pressurizer pressure for an spray are controlling uncomplicated reactor trip l pressurizer pressure is expected to fall within within the P-T limits of this range. Operation of ;
Figure 4-1. pressurizer heaters and spray should be capable of maintaining pressurizer pressure within [2225-2300 psia] z.nd within the Post Accident P-T limits of ,
Figure 4-1.
1 H
SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS Page " of " Revision ""
GUIDELINES SAFETY FUNCTION ACCEPTANCE CRITERIA BASES
- 5. Core Heat Removal a. The RCS loop AT(Ts-Tc) Best estimate analysis Ru i is less than [3*F] showsthat.S&DA will be and less than [3*F] in the
temperature.
Subcooled margin assures adequate core cooling while also accounting for temper-ature variations in the ;
RCS.
- 6. RCS Heat Removal a. At least one steam Adequate RCS heat removal generator has level: will be maintained if at
- 1) within the normal least one steam generator band with feedwater is available for removing available to heat (capable of steam flow
, maintain level and feed flow). The value of of [250 gpm per steam gen-ii) being restored by %erator] is a flowrate based on best estimate analysis.
m;s, sku4 .p y , o,_ werjtWedwater, flow f grnter than [25S [250 gpm per' steam gener-gpr per etarm ator] is sufficient feed querate] w/ Ica/ flow to remove decay heat
, se rec.s in .
and from the core. Decay heat
- b. RCS T,,,is [551-562*F] levels may not be high enough to require a [250 gpm per steam generator]-
feed flowrate. In this i
case, steam generator RT 33 ABB CE SYSTEM 80+*
l i
g SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS Page 5' of 37 Revision ""
GUIDELINES SAFETY FUNCTION ACCEPTANCE CRITERIA BASES
The criteria for a T,,,of
[551-562*F] corresponds.to the control program for steam bypass control valves.
- 7. Containment Isolation a. Containment Pressure [2.0 psig) is based on the less than [2.0 psig] containment pressure alarm.
and It is not expected for an
- b. No containment area uncomplicated reactor trip radiation monitors that containment pressure alarming will increase to the alarm and setpoint.
- c. No steam plant activity monitors alarming During an uncomplicated and reactor trip it is not ex-
- d. No nuclear annex alarms pected that radiation will and be detected inside con-
- e. No reactor building tainment.
al arms.
Steam plant activity is an indication of a'SGTR and is not anticipated for a RT.
- SYSTEM 80 + " TITLE -REACTOR TRIP REC 0VERY EMERGENCY OPERATIONS Page '5 of 37 Revision "" l GUIDELINES l SAFETY FUNCTION ACCEPTANCE CRITERIA BASES .
- 7. Containment Isolation During an un complicated l reactor trip, it is ex-(Continued) pected that Nuclear Annex and Reactor Building alarms will not be received.
- 8. Containment Temperature a. Containment temperature Containment temperature and Pressure Control less than [110*F] less than [110*F] observes and a typical Technical Speci-
- b. Containment pressure fication requirement which less than [2.0 psig]. should not be exceeded for an uncomplicated reactor trip. [2.0 psig] is based on the containment pressure al arm. It is expected that i the containment pressure will not reach this value following an uncomplicated reactor trip.
- 9. Containment Combustible a. Containment temperature Following an uncomplicated -
Gas Control less than [110*F] reactor trip, containment and temperature and pressure l
- b. Containment pressure should not reach [110*F]
less than [2.0 psig]. and [2.0 psig], respec-tively. Maintaining these containment conditions pro-vides an indirect indi-cation that the conditions required for H2generation do not exist. ;
SYSTEM 80 +" TITLE REACTOR TRIP RECOVERY EMERGENCY OPERATIONS l P a g e of 37 Revision ^^"
GUIDELINES Event Strateay ;
This section contains the detailed RT operator actions strategy flowchart, 1
Figure 4-9. The flowchart pictorially depicts the strategy around which the RT guideline is built. It is intended to assist the procedure writer in understanding the intent of the guideline and for use in training. Operators l should understand the major objectives of the guideline in order to permit them to evaluate their progress toward those goals.
The strategy chart shows the recovery guideline strategy in detail and lists the guideline steps which correspond to each strategy objective. Those steps !
which have an asterisk next to the step number can be performed at any time during the event. !
I f
i
I SYSTEM 80 +"- TITLN REACTOR TRIP RECOVERY EMERGENCY OPERATIONS !
GUIDELINES Page' " of " Revision ""
e i
i i
Figure 4-9 I
STRATEGY CHART FOR REACTOR TRIP (FLOW AND STRATEGY CHARTS WILL REFLECT THE DETAILED STEPS IN THE GUIDELINE.)