ML20059H522
| ML20059H522 | |
| 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, NUDOCS 9311100110 | |
| Download: ML20059H522 (21) | |
Text
_
SYSTEM b0 +"
TITLE STANDARD POST TRIP ACTIONS EMERGENCY OPERATIONS p,g,,
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GUIDELINES L
j I
I l
i.
t STANDARD POST TRIP ACTIONS RECOVERY GUIIDELINE
[
t F
SPTA 1
9311100110 931103 C(;;t PDR ADOCK 05200002 A
PDR fj 1
SYSTEM 80 +"
TITLE STANDARD POST TRIP ACTIONS EMERGENCY OPERATIONS GUIDELINES Page of 2'
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2 PURPOSE This guideline provides the immediate operator actions which must be accomplished after an automatic or manually initiated reactor trip. These actions are necessary to ensure that the plant is placed in a stable, safe condition or that the plant is configured to respond to a continuing emergency. This is the entry guideline for the entire Emergency Operations Guidelines (E0G) system. This guideline provides technical information to be used by utilities in developing a plant specific procedure for the System 80+
pl ant.
ENTRY CONDITIONS Any symptom (s) of a Reactor Trip 1.
Any Reactor trip alarms 2.
CEA bottom lights on 3.
Rapid decrease in reactor power 4.
Reactor trip circuit breakers open
'5.
RPS trip setpoint exceeded EXIT CONDITIONS 1.
ALL INSTRUCTION steps of the SPTAs have been performed and all safety function acceptance criteria are verified to be satisfied (implement Reactor Trip Recovery ORG).
9.C 2.
ALL INSTRUCTION steps of the SPTAs have been performed, all necessary CONTINGENCY ACTIONS have been performed, and more than an uncomplicated reactor trip has occurred (implement Diagnostic Actions).
SPTA 2
J
'l SYSTEM 80 +"
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3 INSTRUCTIONS CONTINGENCY ACTIONS
' 1 l
1.
Perform the following to verify
- 1. H any acceptance criteria are j
all safety function acceptance NOT satisfied, Then perform the criteria are satisfied, appropriate contingency actions.
1 REACTIVITY CONTROL
's 1
- 2. Verify reactivity control is
- 2. H reactivity control is NOT l
established by the following:
established, Then do the following as necessary-a.
Reactor power decreasing a.
Manually trip the reactor and
- b. Negative startep rate
- b. Open the reactor trip
- j and breakers c.
Maximum of one CEA not fully
- c. Use Alternate Protection j
inserted.
System to open the CEDM Motor Generator Output Contactor.
t
- d. Deenergize the CEA motor generator l
- e. H more than one CEA NOT i
j fully inserted, Then borate the plant in accordance with Technical Specifications.
L l
SPTA 3
- 4..,
1 SYSTEM 80 +"
TITLE STANDARD POST TRIP ACTIONS
{
t EMERGENCY OPERATIONS Page '
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GUIDELINES i
MAINTENANCE OF VITAL AUXILIARIES l
INSTRUCTIONS CONTINGENCY ACTIONS
- 3. Verify plant electrical power
- 3. Jf electrical power requirements-requirements are satisfied by are NOT satisfied, Then do the ALL of the following:
following as necessary:
a.
Main turbine tripped
- a. Trip the turbine l
- b. Generator output breakers
- b. Open the generator output breakers.
open c.
Non-safety load [13.8 KV] Bus
- c. 'If all Non-safety load [13.8 l
X energized KV) Buses'are de-energized, l
and Then inform the CRS that the i
t Non-safety load [13.8 KV] Bus event is complicated by a j
Y energized loss of power.
j d.
Non-safety load [4.16 KV] Bus d.
If all Non-safety load [4.16 X energized KV] Buses are de-energized, and Then inform the CRS that the Non-safety load [4.16 KV] Bus event is complicated by a l
t Y energized loss of power.
j e.
Permanent Non-safety Load e.
If all Permanent Non-Safety l
[4.16 KV] Bus X energized Load Busses are de-energized, and THEN inform the CRS of the I
Permanent Non-safety Load status of the Permanent Non-
[4.16 KV] Bus Y energized safety Load Buses.
l l
l SPTA 4
f
SYSTEM 80 +"
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5 GUIDELINES i
MAINTENANCE OF VITAL AUXILIARIES (Continued)
(AC & DC POWER) l INSTRUCTIONS CONTINGENCY ACTIONS l
- f. Safety Load Division I f i)
If Safety Load Division I l
energized via Permanent Non-energized Safety Bus X and and Safety Load Division II Safety Load Division II energized,THEN note which energized via Permanent Non-divisions are powered by Safety Bus Y Emergency Diesel Generators.
ii)
If Safety Load Division I NOT energized f
and Safety Load Division II NOT energized THEN energize appropriate Safety Load Division from the Altern.ite AC Power l
Supply.
i iii)
If EITHER Safety Load j
Divisions NOT energized l
4 THEN initiate actions to investigate cause of de-energization to the f
affected Safety Load j
Divisions.
i SPTA 5
f
SYSTEM 80 +"
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GUIDELINES MAINTENANCE OF VITAL AUXILIARIES (Continued)
INSTRUCTIONS CONTINGENCY ACTIONS g.
- g. i) H [125V] DC and [120V]
Safety Bus A energized AC Safety Bus A NOT and energized and
[125V) DC and [120VJ AC Safety Bus B energized Safety Bus C NOT and energized
[125V] DC and [120V] AC OR Safety Bus C energized
[125V) DC and [120V] AC and Safety Bus B NOT
[125V) DC and [120V] AC energized Safety Bus D energized and (125V) DC and [120V] AC Safety Bus D NOT energized THEN notify CRS of the loss of a DC train.
SPTA 6
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SYSTEM 80 +"
TITLE
.cVANDARD POST TRIP ACTIONS EMERGENCY OPERATIONS GUIDELINES Page _'_ of "
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RCS INVENTORY CONTROL INSTRUCTIONS CONTINGENCY ACTIONS 4.
Verify RCS inventory control is
- 4. H RCS inventory control is NOT established by the following:
established, Then do the following as necessary:
a.
Pressurizer level:
a.
Verify proper operation of i)
[2%to 78%]
the PLCS.
and
- b. Take manual control of the ii) trending [33I.to 52%]
PLCS to restore and maintain and pressurizer level [33%to 52%)
- b. The RCS is subcooled.
- 5. Verify RCS pressure contrc1 is
- 5. H RCS pressure control is NOT established by pressurizer established, Then do the 2
pressure:
following as necessary:
a.
[2160 to 2370 psia]
a.
Verify proper operation of i
and the PPCS b.
trending to [2225 to 2300
- b. Take manual control of the psia]
PPCS to restore and maintain j
pressurizer pressure [2225 to 2300 psia]
l
- c. H pressurizer pressure de-f creases to [1825 psia], Then i
ensure an SIAS is initiated
- d. H pressurizer pressure de-j creases to less than [1400 psia] following a SIAS, Then trip 2 RCPs (in opposite loops).
SPTA 7
SYSTEM 80+"
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GUIDELINES CORE HEAT REMOVAL l
INSTRUCTIONS CONTINGENCY ACTIONS
~
t-
- 6. Verify core heat removal via
- 6. Jf forced circulation core heat i
forced circulation by the removal is NOT possible, Then following:
verify natural circulation is developing by:
{
a.
At least one RCP is operating
- a. Loop aT (Tn - T,) i' less and than normal full power AT,
- b. T - T, is less than [3*F]
- b. Cold leg temperatures n
and constant or decreasing, j
a c.
The RCS is subcooled.
- c. The RCS is subcooled, i
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t 4
i 4
SPTA 8
I SYSTEM 80 +"
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GUIDELINES RCS HEAT REMOVAL j
i INSTRUCTIONS CONTINGENCY ACTIONS 7.
Verify RC: neat removal by the 7.
- a. H RCS T,,, > [562*F], Then do
[
following:
the following:
a.
At least one SG has level:
i)
Ensure main, startup or-i) within the normal emergency feed is con-level band with trolling or restoring maintain level level to at least one SG.
[
or ii)
Ensure the turbine bypass j
ii) being restored by a system is operating to i
Main, Startup or control RCS T,y, [551 to Emergency feedwater 562*F].
j i
flow iii) H turbine bypass system NOT available, Then use and ADVs to control T,,, [551
{
I to 562*F]
b.
RCS T,,, [551 to 562*F]
- b. H RCS T,,, < [551*F] then do the following:
{
i) Ensure feed flow not ex-and cessive. Control main, startup or emergency feed j
c.
SG pressure [1050 to 1150 flow as necessary.
psia]
ii)
H SG pressure < [1000 l
psia], Then ensure tur-bine bypass valves, ADVs, l
t and MSSVs are closed.
{
iii)
H SG pressure s [843 l
psia), Then ensure MSIS activated and consider i
ESDE.
l l
l SPTA 9
SYSTEM 80 +"
TITLE STANDARD POST TRIP ACTIONS EMERGENCY OPERATIONS P a g e '"
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j GUIDELINES CONTAINMENT ISOLATION INSTRUCTIONS CONTINGENCY ACTIONS
}
8.
Verify normal containment
- 8. Jf normal containment environment by:
environment is NOT indicated, Then do the following as l
necessary:
i a.
Containment pressure less
- a. J_f containment pressure is than [2 psig]
greater than or equal to [2.7 and psig], Then ensure a CIAS b.
No containment area radiation b.
If Nuclear Annex radiation monitors alarming levels are at or above the j
and alarm setpoint, Then consider i
I a LOCA outside containment.
c.
No Nuclear Annex radiation c.
If Reactor Building radiation alarms.
levels are at or above the j
and alarm setpoint, Then consider
}
a LOCA outside containment.
d.
No Reactor Building radiation
- d. J1 Steam Plant radiation al arms.
levels are at or above the and alarm setpoint, Then consider f
SGTR.
c e.
No Steam Plant radiation l
al arms.
.j i
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SYSTEM 80 +"
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GUIDELINES CONTAINMENT TEMPERATURE & PRESSURE CONTROL INSTRUCTIONS CONTINGENCY ACTIONS 9.
Verify normal containment
- 9. H normal containment l
temperature and pressure temperature and pressure parameters by the following:
parameters are NOT indicated, Then do the following as necessary:
a.
Containment temperature less
- a. Verify All available than [110'F]
containment fan coolers i
and operating t
9.E b.
Containment pressure less
- b. H containment pressure than [2 psig]
greater than or equal to [8.5 psig], Then ensure at least one containment spray header is delivering at least [5000 gpm]
[
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I m
.m
SYSTEM 80 + =
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CONTAINMENT COMBUSTIBLE GAS CONTROL 4
INSTRUCTIONS CONTINGENCY ACTIONS 10.
Verify no expected increase 10.
E normal containment-in containment combustible temperature and pressure i
gas concentration by the parameters are NOT indicated, following:
Then do the following as necessary:
l a.
Containment temperature less
- a. Verify proper functioning of than [110*F]
all available containment and recirculation cooling units b.
Containment pressure less
- b. Verify proper functioning of than [2 psig) all available containment air recirculation fans, including the pressurizer compartment, reactor compartment, and CEDM cooling units, if they are available.
11 -.
Verify all safety function 11.
H all safety function acceptance criteria are acceptance criteria are NOT satisfied and the event is an satisfied, Then more than a j
uncomplicated reactor trip, uncomplicated reactor trip Then implement the Reactor has occurred and the Trip Recovery Guideline.
Diagnostic Actions guideline must be implemented.
SYSTEM 80 +"
TITLE STANDARD POST TRIP ACTIONS EMERGENCY OPERATIONS Page " of 2' - Revision ""
GUIDELINES Bases for Operater Actions i
This is the entry guideline for the SYSTEM 80+ E0G System. This guideline is used for any event which actuates or requires a reactor trip.
It is intended j
that the operator check each safety function and perform the contingency actions if necessary. The acceptance criteria are selected from best estimate 1
analysis to reflect the range for each parameter which would be expected
[
following a relatively uncomplicated reactor trip. The recovery actions are selected to reflect the need to verify the actuation of automatic systems and i
to perform appropriate post trip actions which will ready the plant to respond to any event, p
1.
Standard Post Trip Actions (SPTAs) are organized around those critical safety functions which must be satisfied when a reactor trip is actuated or required, in order to ensure that the plant is placed in a stable, safe condition or that the plant is configured to further respond to a continuing casualty.
In order to provide for this, the operator is given specific, unambiguous acceptance criteria which can be evaluated l
without interpolation directly from the control room instruments. These j
acceptance criteria are located under the " INSTRUCTIONS" heading. These criteria (and the range of the numerical criteria) are chosen from best i
estimate analyses to bound the expected conditions which would follow a j
relatively uncomplicated reactor trip. Thus, checking the acceptance l
criteria serves two purposes: if the acceptance criteria are met, this j
serves as a verification that the safety function is being fulfilled; second, meeting the acceptance criteria is a diagnostic indicator (i.e.,
t meeting all of the acceptance criteria implies that nothing more serious l
1 f
than a relatively uncomplicated reactor trip has occurred).
If the acceptance criteria are not met, then this serves as a cue to perform l
the appropriate contingency actions located under the heading of the j
same name. These actions are chosen to reflect the verification of expected automatic system responses and the usual, easy to accomplish I
4 actions which operators always take in response to a trip.
l
SYSTEM 80 +"
TITLE STANDARD POST TRIP ACTIONS f
EMERGENCY OPERATIONS GUIDELINES Page "
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2.
The Reactivity Control safety function is designed to ensure that the i
reactor is shutdown in order to reduce heat input to the RCS. The Reactivity Control acceptance criteria are chosen to reflect those j
reactor conditions which would prevail during the first ten minutes i
following a trip. No more than one CEA stuck out is chosen as the
]
cutoff point in the third criterion since it is a core design criterion that the reactor will be shutdown even with the most reactive rod stuck out. Contingency actions a) through d) are directed at inserting the CEAs.
Contingency action e) reflects the Technical Specification f
requirement to borate the RCS shculd more than one CEA not be fully inserted.
i 3.
Maintenance of Vital Auxiliaries is chosen as the next safety function j
to address since the electrical system is essential to the continued fulfillment of succeeding safety functions.
The acceptance criteria j
reflect the automatic disconnect of the main turbine generator and the transfer of power to offsite which should occur immediately upon a trip.
l Contingency actions are chosen to remedy the failure of automatic system responses and.to ensure that the emergency diesel generators and the
{
alternate AC source are available to supply AC power, if necessary.
I i
i 4.
RCS Inventory Control and RCS Pressure Control are next in order of priority due to their importance to core cooling and their potential for rapid change.
RCS Inventory Control is intended to ensure an adequate amount of fluid is in a subcooled state to remove decay heat. During a relatively uncomplicated reactor trip, the pressurizer will retain an indicated level between [2 to 78%) (even though the pressurizer heaters i
may be deenergized briefly on low pressurizer level) which is acknowledged in the acceptance criteria. The upper limit of pressurizer level is based on avoiding solid water operations. The lower limit is based on having some water in the pressurizer.
If the pressurizer level control system functions properly, level in the pressurizer should be trending to [33 to 52%1. Contingency actions are selected to observe SPTA 14 ABB CE SYSTEM 80+*
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l SYSTEM 80 +"
TITLE STANDARD POST TRIP ACTIONS l
i EMERGENCY OPERATIONS Page
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proper operation of charging and letdown by the PLCS.
Failing that, l
manual control of the PLCS or the charging and letdown is taken to
^
j restore and maintain pressurizer level [33 to 52%].
l 5.
RCS Pressure Control relates to the maintenance of RCS fluid in a l
subcooled condition in order to adequately remove decay heat. Best estimate analyses reveal that for a relatively uncomplicated reactor trip, pressure will remain within the range of [2160 to 2370 psia]. The l
limits are adequate to ensure adequate RCS subcooling and to prevent the lifting of primary safety valve.
If the pressurizer pressure control j
system functions properly and pressurizer level is above the heater cutoff setpoint, pressurizer pressure should be trending to between
[2225 and 2300 psia]. Contingency actions are directed at observing proper operation of pressurizer heaters and spray by the PPCS.
Failing that, actions are directed at restoring or maintaining pressure [2225 to 2300 psia] with manual control of the PPCS or the pressurizer heaters l
and spray.
If pressurizer pressure decreases to or below the SIAS
]
setpoint, SIAS should be initiated automatically.
If this does not occur, the operator should manually initiate SIAS. While performing the 2
i Standard Post Trip Actions, the operator is instructed to trip two RCPs i
(in opposite loops) if pressurizer pressure decreases to less than [1400 psia] following SIAS. A SIAS is specified to distinguish between a
]
controlled and an uncontrolled depressurization.
If two RCPs are tripped early in the event, the plant can be maintained in a safe i
condition regardless of event diagnosis. This action provides the
.j operator with maximum flexibility for plant control while still ensuring a conservative approach to event recovery.
6.
Core Heat Removal is related to circulating cooling fluid in a subcooled state through the core to remove decay heat. The acceptance criteria assume RCPs are running (as they would be following a relatively uncomplicated trip) thereby providing the small loop AT [<3*F] expected with decay heat.
Subcooling is concerned with maintaining adequate SPTA 15 ABB CE SYSTEM 80+"
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GUIDELINES F
fluid conditions surrounding the core. Contingency actions are directed at removing heat via the steam generators using subcooled natural circulation.
If all RCP operation is terminated and inventory and pressure are controlled, then natural circulation is monitored by heat removal via at least one steam generator. Natural circulation flow should occur within 5-15 minutes after the RCPs are tripped and, thus, may not be
[
established during the time frame it which the SPTAs are being performed.
i Natural circulation is governed by decay heat, component elevations, primary to secondary heat transfer, loop flow resistance, and voiding.
Component elevations on the System 80+ plant are such that satisfactory natural circulation decay heat removal is obtained by fluid density differences between the core region and the steam generator tubes.
The operator has adequate instrumentation to monitor natural circulation for the single phase liquid natural circulation process.
The RCS temperature instrumentation, namely loop AT, can be used along with other information to confirm that the single phase natural circulation process is effective. The natural circulation process involving two phase cooling is complex and varied enough so that RCS loop AT may not he a meaningful indication of adequate natural circulation cooling. The i
guidelines are written to alert the operator to use explicit acceptance criteria for natural circulation only when RCS inventory and pressure i
are controlled.
l l
The RCS temperature response during natural circulation will usually be slow 5-15 minutes as compared to a normal forced flow system respon;e i
time of 6-12 seconds, since the coolant loop cycle time will be significantly longer.
.-e
SYSTEM 80 +"
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f When single phase circulation is established in at least one loop, the j
RCS indicates all of the following conditions:
t a.
Loop aT (Tn - T ) less than normal full power 4T, c
b.
Cold leg temperatures constant or decreasing, l
c.
RCS at least subcooled (verifies single phase flow),
d.
No abnormal differences between T RTDs and core exit t
n thermocouples. Hot leg RTD temperature should be consistent with the core exit thermocouples. Adequate natural circulation flow ensures that core exit thermocouple temperatures will be approximately equal to the hot leg RTD temperatures.
If the criteria listed above are not satisfied, then the operators should ensure that RCS pressure and inventory, and SG steaming and feeding, are being controlled properly.
7.
RCS Heat Removal is next in priority because the parameters associated with it are concerned mostly with steam generators, which are the primary means of removing heat from the RCS.
Furthermore, steam l
generator level and pressure also have the potential for rapid change.
Instruction step a) is to ensure the presence of an operable steam generator for removing heat. The steam generator level may briefly transit below the narrow range steam generator level indication.
RCS average loop temperature (criterion b) in the range of-[551 to 562*F) is indicative that steam generators are adequately removing heat.
i Instruction Step c) also ensures an operable steam generator for r
controlled removal of heat. The steam generator pressure range given i
provides the expected range maintained by the steam bypass control system.
The upper steam generator pressure limit, [1150 psia] is below the MSSVs setpoint and the lower limit, [1050 psia] would be indicative of an excessive cooldown. The contingency actions relate to feed and or g
steam flow to the steam generator under one of two conditions:
t
SYSTEM 80 + =
TITLE STANDARD POST TRIP ACTIONS EMERGENCY OPERATIONS p
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j GUIDELINES a.
RCS heat removal is NOT sufficient (RCS T.,, > [562*F].
l or b.
RCS heat removal is excessive (RCS T,,, < [551*F]
If RCS Heat removal is not sufficient (e.g., RCS T,., > [562*F] due to
+
loss of condenser vacuum), the operator is provided several items to check in order to re-establish RCS heat removal.
Feedwater must be supplied to at least one steam generator in order to ensure adequate heat removal will be maintained.
If the turbine bypass system is not functioning properly in automatic, the operator should attempt to take manual control to restore RCS T,,, to [551 to 562*F].
If manual control of the turbine bypass system is not possible or condenser vacuum is lost, then the atmospheric dump valves are operated to control T,,,
between [551 and 562*F].
If RCS heat removal is excessive (e.g., RCS T,,, < [551*F] due to stuck open main steam safety valves), then guidance is provided on how to mitigate this transient. The operator should ensure that the feed rate to the steam generators is not excessive.
Because decay heat and power history will vary over core life, the operator must use judgement in feeding the steam generators.
If the refill rate is too fast, RCS temperature can easily be driven below the desired no load value.
If an overfeed condition is not corrected, pressurizer level may decrease to the point where the pressurizer is drained and the safety injection j
system is actuated.
Following a reactor trip, steam generator pressure l
should increase to approximately [1150 psia), and a band of [1050 to s
1150 psia] is specified in the acceptance criteria.
If steam generator j
pressure decreases to less than [1000 psia], then some system abnormality exists that should be investigated and corrected. The limit of [1000 psia] was chosen because it is far enough below the turbine SPTA 18 ABB CE SYSTEM 80+"
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GUIDELINES bypass system control program to minimize unnecessary operator actions while still high enough to give the operator time to find and correct the problem prior to a main steam isolation signal (MSIS). Possible sources of the excessive heat removal (turbine bypass valves, ADVs,
{
MSSVs) should be checked shut or manually closed, if possible.
If steam l
generator pressure decreases to or below [843 psia], then an MSIS should be actuated. The operator should ensure this has occurred, taking manual actions as necessary, and consider an ESDE when performing Diagnostic Actions.
8.
Containment Isolation serves to ensure that radionuclides are contained inside the containment building. The acceptance criteria are designed j
to ensure that a normal containment environment exists. High I
containment pressure, (above the alarm setpoint [2 psig]) or radiation alarms in the and steam plant, the Nuclear Annex and/or the Reactor Building are indications that more than a relatively uncomplicated reactor trip has occurred. Contingency actions are designed to ensure that the containment is isolated when necessary (CIAS occurs when containment pressure is greater than or equal to [2.7 psig]) and that a 4
SGTR is considered when performing Diagnostic Actions if turbine f
building radiation alarms are obtained.
f 9.
Containment Temperature and Pressure Control has as its goal the preservation of the containment building boundary by preventing or l
minimizing pressure excursions.
Since containment temperature and j
pressure are not expected to change noticeably for a relatively l
uncomplicated reactor trip, the acceptance criteria are selected to be sensitive to any change. Contingency actions focus on restoring or l
initiating containment cooling either with the containment cooling fans, with the containment spray system, or with a combination of these two systems.
If containment temperature and/or pressure have exceeded their f
expected values but containment pressure is less than [8.5 psig], then
[
i the operator should verify that ALL available containment fan coolers l
{
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GUIDELINES t
are operating.
If containment fan cooler operation is not adequate to maintain containment pressure less than [8.5 psig], then the containment spray system should be actuated automatically.
If containment pressure reaches [8.5 psig], then the operator should verify that containment j
sprays have automatically actuated.
If this has not occurred automatically, then the operator should take necessary steps to manually actuate containment spray.
In order to maintain containment pressure l
below design pressure in the event of a design basis event, there exist l
redundant containment spray trains for containment heat removal.
10.
In the Standard Post Trip Actions, checking Containment Combustible Gas Control serves to alert the operator of the potential for hydrogen generation. Hydrogen can be generated (and released to the containment) by several mechanisms:
metal-water reactions involving Zirconium or stainless steel in a.
the RCS, b.
corrosion reactions involving the containment spray solution and I
metals (Zinc and aluminum) inside containment, radiolytic decomposition of water due to fission product decay.
c.
i The metal-water reactions and the radiolysis are of concern during LOCAs when the hydrogen generated by these mechanisms can escape to the l
containment atmosphere (through the RCS break).
The corrosion reactions l
require high temperatures to produce significant amounts of hydrogen.
[
This mechanism is of concern during LOCAs and steam line breaks inside l
r containment, since these events may result in high temperatures and
{
containment spray actuation.
The potential for hydrogen generation is identified by increasing containment pressure and temperature, since these indicate that a LOCA l
P
SYSTEM 80 +"
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or steam line break may be occurring inside containment and may result j
in containment spray actuation. Contingency actions are intended to minimize the amount of hydrogen generated due to corrosion reactions (by reducing the containment temperature), and to prevent local accumulations of hydrogen (by mixing the containment atmosphere with all I
available fans).
i 11.
If all safety function acceptance criteria (located under the j
INSTRUCTIONS heading) are verified to be satisfied and the operator has determined that nothing more than a relatively uncomplicated reactor trip has occurred, the Reactor Trip Recovery Guideline should be j
implemented. This guideline will provide the guidance necessary to i
place the plant in a stable, safe condition, to perform an RCS cooldown, l
or to prepare for a possible reactor startup.
If all acceptance criteria are not satisfied, then more than a uncomplicated reactor trip has occurred. While performing the Standard Post Trip Actions the operator will be obtaining diagnostic information.
The E0G Diagnostic Actions guideline must be utilized to obtain event diagnosis.
i j
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i R
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