ML20059H582
| ML20059H582 | |
| 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-931103-02, NUDOCS 9311100130 | |
| Download: ML20059H582 (19) | |
Text
I' SYSTEM '80 + =
TITLc DIAGNOSTfC ACT10NS l
w EMERGENCY OPERAT!ONS GUIDELINES page of
Revision "
i
.i k
r DIAGNOSTIC ACTIONS i
r
'i I
it DA I
9311100130 931103 P'
PDR ADOCK 05200002 b
A PDR-
[i
SYSTEM 80 +"
TITLE DIAGNOSTIC ACTIONS l
EMERGENCY OPERATIONS Page of " Revision ""
2 GUIDELINES PURPOSE This guideline provides the Diagnostic Actions which must be performed to attempt to determine a preliminary diagnosis of the event (s) which has resulted in a l
These actions are performed to best enable the operator to j
determine whether to implement an ORG or the FRG.
This guideline contains no specific operator action steps but does provide technical information to be used by utilities in developing a plant specific diagnostic procedure.
l ENTRY CONDITIONS i
The Standard Post Trip Actions have been performed.
EXIT CONDITIONS i
The Diagnostic Actions have been performed ar.d the preliminary diagnosis l
l indicates one of the following:
The diagnosis of one event is apparent and an appropriate Optimal a.
Recovery Guideline can be implemented to mitigate the event.
l l
t' b.
The diagnosis of one event for which emergency guidance exists is not possible and the Functional Recovery Guideline must be I
implemented.
1 i
l t
t i
l i
i DA 2
SYSTEM 80 +"
TITLE DIAGNOSTIC ACTIONS EMERGENCY OPERATIONS GUIDELINES Page of "
Revision "
3 FIGURE 1 DIAGNOSTIC ACTIONS POWER OPERATION 1r EVENT r
- 1 YES REACTOR TRIP No (ACTUATED OR REQUIRED)
U PERFORM STANDARD POST TRIP ACTIONS (SPTAs) 1r l
- 2 ARE ALL MA YES NO ACCEPTANCE CRITER!A MET 7 r
y GO TO REACTOR TRIP RECOVERY Y
DA 3
SYSTEM 80+"
TITLE-
- DIAGNOSTIC ACTIONS EMERGENCY OPERATIONS Page '
of "
Revision a"'
GUIDELINES FIGURE 2.
DIAGNOSTIC ACTIONS 1
V I I f
O Ans MEAcTMTT YES cosmoot go REcuumEneEstTs/
If II
/
4 aoTo
- 4 l
AT RECOVERY i
LEAsf1 oumsunt yES NO sa,an om,.
oF Ac power ENEAQQED H
r 7
If jr
- 5 d,
YES AT, NO l
YES SAPETYoMesou No c
nuwesNQ oF oC POWEA f
?
1I cosecen r
coweman toss STA1 JN oF oPrgrrg NT poWEA EVENT i
1T
~
i 1r AT "d **
YES NO AoEouATE l
PEED 1I i
consetstoes or rEsowArun i
=
1r 2
V i
DA 4
f SYSTEM 80+"
TITLE orasN0STn, ACTIONS EMERGENCY OPERATIONS Page,
of,,
Rev..ision _
GUIDELINES FIGURE 3 DIAGNOSTIC ACTIONS r
\\/
v
- 8 YES NO 9
L 1r 1r CONSIDER EXCESS pg STEAM DEMAND gg EVENT ONTAINM YES PRESSURE
>[2.0 PStG] OR MCREASWQ 7
1r if CONSIDER LOSS OF
- 10 COOLANT EVENT YES ACTMTY N NO STEAM PLANT 7
- 11 i
NO YES OR om eu j
ALAAass e
If 1 r gag lgrJa$$
CON.OER STEAu GENERATOR TUSE UN RUPTURE EVENT i
t I
)I l I i r 3
- REFEA TO WEAK
(/
- 9tSERT *SUSC00UNG 98CREAS8P3*
QB 43 MES$ufE ABNOMILALLY LOW v
IDDmFEATON CHART FK3URE 59 DA 5
SYSTEM 80 + "
TITLE DIAGNOSTIC ACT10NS EMERGENCY OPERATIONS Page '
of " Revision ""
GUIDELINES FIGURE 4 DIAGNOSTIC ACTIONS 3
\\/
1r
- 12 DMGNOSIS OF NO YES ONE EVENT APPARENT 7 1f 1 I GO GO TO APPROPRIATE gN OML OPTIMAL RECOVERY RECOVERY GUIDELINE GUIDLINE When the Diagnostic Actions are complete, a preliminary diagnosis should be apparent with an appropriate ORG available to be implemented (Reactor Trip Recovery is an ORG), or a preliminary diagnosis of one event is not possible and the FRG must be implemented.
EBQ 1
DA 6
4 l
SYSTEM 80 +"
TITLE-DIAGNOSTIC ACTIONS t
EMERGENCY OPERATIONS 7
GUIDELINES Page of " Revision ""
i SUPPLEMENTARY INFORMATION i
I This section contains items which should be considered when implementing EOGs and
' i preparing plant specific E0Ps. The items should be implemented as precautions, l
cautions, notes, or in the E0P training program.
1.
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).
2.
Instrumentation required to provide key parameters for event diagnosis
(
early in an event should be obtained for the most reliable sources, e.g.
DIAS-P.
i r
t i
i i
b l
DA 7
SYSTEM 80 +"
TITLE DIAGNOSTIC ACTIONS EMERGENCY OPERATIONS Page of " Revision ""
8 GUIDELINES FIGURE 3-1 BREAK IDENTIFICATION CHART PRIMARY BREAK OR SECONDARY BREAK SUSPECTED
YES NO o
my K p M M MSE l
OF sMALL BREAK LOCA N cONTAmuENT 1r ExcEsssTEAu DEMAND EVENT not)
If PRIMARY saoE mREAx 1r 1r B.
Yts CONT
~
NO YES MMNMENT gg PRESSURE PRESSURE acREAsmo ocREAsm ir D.
Y NO Yts STEAM PLANT 1r 1r lf II I f
SYSTEM 80+"
TITLE DIAGNOSTIC ACTIONS EMERGENCY OPERATIONS Page '
of "
Revision "
GUIDELINES BASES For the purpose of the bases explanations, each question block on the chart has a unique identification number.
These numbers would not be necessary for the actual plant specific E0P Diagnostic Flowchart (except in the bases section if desired).
Block #1 If an abnormal condition occurs while the reactor is in MODE 1 or 2, as long as a reactor trip does not occur or is not required, plant operation can continue.
The plant Abnormal Operating Procedures (A0Ps) would most likely be the governing documents.
Once a plant trip occurs or is required, the Standard Post Trip Actions (SPTAs) are implemented.
Block #2 By the time the Diagnostic Actions are implemented, the operator has already completed the Standard Post Trip Actions (SPTA) in response to the reactor trip and any other concurrent plant component or system failures. The operator has already made an initial evaluation of plant status and because the Standard Post Trip Actions also constitute a check of the safety functions, the operator is i
I also aware of the status of safety functions.
If no safety functions were in jeopardy, that is, all of the safety functions met their. respective acceptance criteria (only instruction steps performed), then nothing more than an uncomplicated reactor trip has occurred. If one nr more safety functions did not meet the acceptance criteria of the SPTA (one or more contingency actions
]
I performed), the operator must attempt to determine a preliminary diagnosis of the event (s) which has resulted in a reactor trip.
The Diagnostic Actions Guideline has been developed to assis; the operator in logically deciding whether to implement an appropriate ORG or to implement the l
FRG. Minor system failures will not impair the use of this diagnostic flowchart a
l DA 9
I SYSTEM 80 +"
TITLE DIAGNOSTIC ACTIONS EMERGENCY OPERATIONS Page '
of "
Revision "
GUIDELINES c
in distinguishing symptom sets.
To go much beyond the complexity of this diagnostic flowchart will require too much operator time and may hinder the operator in performing the required actions in a timely manner.
For depressurization events, the operator can also refer to the Break Identification l
Chart (Figure 3-1) to aid in diagnosing the event. As noted in these diagnostic aids, the trending of key parameters provides valuable information which should l
be utilized in determining the preliminary diagnosis.
A particular ORG would be implemented after the operator has completed t. e Standard Post Trip Actions and has been able to diagnose one event taking place.
Certain events (i.e, LOCA, SGTR, ESDE, and LOAF) do not require offsite power in order to adequately mitigate the effects of the accident.
For this reason the LOCA, SGTR, ESDE, or LOAF ORG is to be implemented even if a loss of Offsite j
Power has also occurred. The ORG will mitigate the event and maintain the health and safety of the public.
The ORGs provide a continual verification that the preliminary diagnosis was correct.
In essence, this diagnostic confirmation process is provided by checking the status of safety functions using the ORG i
Safety Function Status Checks.
If the ORG in use continues to satisfy SFSC acceptance criteria, the ORG guidance is followed.
If the ORG in use is inadequate, either because new information on symptoms appears that is not
[
covered in the ORG, or because the SFSC acceptance criteria are not being j
satisfied, then a transfer is made to the appropriate ORG. If identification of a particular symptom set is not possible, the operator implements the Functional j
Recovery Guideline.
l Figure 3-2 provides a basic flowchart showing how the Diagnostic Actions and the ORG diagnostic confirmation process are utilized to ensure that the proper guideline is implemented to mitigate the event (s) in progress.
Since the flowchart is attempting to diagnose a specific event through the investigation of a series of symptoms, it is not arranged in the order of the safety function hierarchy. Instead, the integrated plant and integrated safety functions have been considered, and the symptoms to be investigated have been DA 10 ABB CE SYSTEM 80+'
SYSTEM 80 +"
TITLE DIAGNOSTfC ACTIONS EMERGENCY OPERATIONS l
Page "
of " Revision ""
GUIDELINES prioritized so that a determination of the event in progress can be made in the i
most efficient manner.
The operator should implement the entire Diagnostic Flowchart to determine if a single event is in progress, or if there are multiple events in progress. There are two exceptions to this guidance.
If Reactivity Control has not yet been established, or if a complete loss of station power has occurred, the operator should proceed directly to the Functional Recovery Guideline (FRG).
These exceptions are necessary because no ORG will be successful unless the reactor is shut down, or if there is no vital AC or DC l
power available.
When evaluating parameters and to answer questions contained in the Diagnostic Flowchart, the operator should consider past and present trends such that all available information is used to decide which event is in progress. For example, f
if a Steam Generator Tube Rupture (SGTR) event occurs, it is possible that the main steam line radiation monitors will alarm prior to the reactor trip due to f
the N-16 present in the primary system leaking into the Steam Generators.
However, after the reactor trips, the alarms may clear (unless latched) due to the significantly lower production of N-16.
In this case, the operator should still consider the fact that the main steam line radiation monitors had alarmed 3
(and latched) prior to the trip (even though they are not currently in alarm).
When this information is combined with the N-16 monitor information, the operator will answer "YES" to block #10.
That way, the operator will be appropriately directed to consider a Steam Generator Tube Rupture event.
i Block #3 i
If the reactivity control requirements are not met, the operator should proceed i
directly to the Functional Recovery Guideline, since no ORG will be successful unless the reactor is shut down. Therefore, if the operator answers "N0" to the Reactivity Control question, the next b'ock on the flowchart says "G0 TO FUNCTIONAL REC 0VERY GUIDELINE" instead of " Consider" Functional Recovery l
Guideline.
l l
4
SYSTEM 80+"
TITLE DIAGNOSTIC ACTIONS
^
EMERGENCY OPERATIONS GUIDELINES Page
'2 of "
Revision "^"
Block #4 If the plant safety equipment has no electrical power, cooling water, j
ventilation, etc., the steps in the ORGs will not be effective. Therefore, the status of the Vital Auxiliaries is considered next.
}
If at least one safety division of AC power is available, the operator proceeds to the next question on the flowchart (Block #5), since all of the ORGs are written to accommodate the availability of only one safety division of AC power.
If there is no Vital AC power, the event in progress could be a Station Blackout, a combination of a Station Blackout and another event, or a complete loss of station power. The operator is directed to Block #6 to consider the status of DC power.
2 Block #5 r
The operator is asked if at least one RCP is running.
There could be several reasons why all four RCPs are not running (Loss of Offsite Power, Loss of CCW to the RCPs, electrical faults on the 13.B KV buses, etc.). The question is asked this way to determine if a " loss of flow" event has occurred either from a loss Of Offsite Power (LOOP), or for any other reason.
If it has, this would not be considered an uncomplicated reactor trip and the Reactor T. rip Recovery ORG would not be appropriate because the RTR ORG assumes forced RCS flow conditions exist.
The Loss Of Offsite Power (LOOP) ORG is written for either a LOOP or a Loss of Forced Circulation, since LOOP is essentially a maintenance of natural circulation event coupled with actions to restore offsite power to the station if applicable.
It is implemented on'.v if the Loss of Offsite Power / Loss.of Forced Circulation event is the only event in progress. Therefore, for the Loss-of Offsite Power / Loss of Forced Circulation event, the operator is instructed to
" Consider" the LOOP and proceed to the next block on the flowchart to determine if other events are also in progress.
SYSTEM 80 +"
TITLE DIAGNOSTIC ACTIONS EMERGENCY OPERATIONS Page of,,
Rev..ision.
o GUIDELINES Block #6 i
If there is no station power (i.e., no vital AC and no vital DC), the operator is directed to the FRG since no ORG will be effective.
If at least one safety i
division of DC power is available and not other event is in progress, the Station Blackout (SB0) ORG should be implemented.
However, since only the Reactivity l
Control and Vital Auxiliaries safety functions have been investigated thus far, the Diagnostic Flowchart instructs the operator to " Consider" 580. The operator will then continue to proceed through the rest of the Diagnostic flowchart to determine if only an SB0 is in progress, or if it is combined with some other event.
Block #7 I
The operator is asked if at least one Steam Generator (SG) has adequate feed (as defined by the SPTA steps).
If the answer is "N0", the operator is instructed i
to " Consider" the loss of All Feedwater (LOAF) event, and proceed to the next block on the flowchart.
Otherwise, SG feeding is considered adequate and the operator proceeds directly to the next block on the flowchart.
f Block #8 The purpose of this block is to help the operator distinguish a loss Of Coolant l
Accident (LOCA) or Steam Generator Tube Rupture (SGTR), from an Excess Steam Demand Event (ESDE). When the plant specific E0Ps are developed, the plant has the option of either_ placing the question "IS SUBC00 LING INCREASING?", or the question "IS SG PRESSURE ABNORMALLY LOW 7", or both questions in this block. If l
a LOCA or SGTR is in progress, it is not expected that subcooling would be
'{
increasing (unless it is a small break of a size that allows the SI system to refill the Pressurizer, thereby causing Pressurizer pressure to increase).
If the event in progress is an ESDE, subcooling should be increasing (unless the l
affected SG has already blown dry). In either case the operator should consider l
past and present trends to determine which event is in progress. This block will DA 13 ABB CE SYSTEM 80+*
i 8
SYSTEM 80 +"
TITLE DIAGNOSTIC ACTIONS 1
EMERGENCY OPERATIONS GUIDELINES Page "
of "
Revision "
j assume that at least during the initial stage of the LOCA or SGTR, the break is large enough to cause subcooling to be constant, or to decrease. In this case, the Diagnostic Flowchart will direct the operator to block #9. This block will also assume that at least during the initial stage (blowdown phase) of the ESDE, subcooling will increase and/or pressure will be abnormally low in the affected SG. The operator will be instructed to consider an ESDE.
Block #9 The purpose of this block is to help the operator distinguish between a LOCA i
inside containment, and a LOCA outside of containment or an SGTR. If containment i
pressure is increasing, it is assumed that a LOCA exists inside of containment and the operator is directed to consider the LOCA event. If containment pressure is not increasing, the operator is directed to block #10.
t Block #10 e
The purpose of this block is to help the operator distinguish between and SGTR or a LOCA outside of containment.
If there is activity indicated in the steam plant, a Steam Generator Tube Rupture should be considered. Alarms are not used as the sole means of determining whether or not and SGTR event is occurring because the activity levels in the secondary system may not yet be high enough to trip the alarm setpoints.
Alarms were used in SPTA to get a quick " big picture" of the plant status (SPTA is not intended to be used for diagnosing the event, although some events are suggested based upon fairly definitive indications).
If the steam plant activity alarms were energized in SPTA, it could be assumed with reasonable certainty that at least an SGTR event was occurring. However, this Diagnostic Flowchart takes a closer look at the plant and its indications for the purpose of identifying the actual event (s) in progress. If there is no significant activity in the steam plant (and activity j
levels are not increas,ing), an SGTR is not assumed to exist and the operator is
]
directed to block #11.
s SYSTEM 80+"
TITLE D! AGNOSTIC ACTIONS EMERGENCY OPERATIONS Page
'5 of " Revision ""
GUIDELINES I
Block #11 i
i This block is used to determine if a LOCA exists outside of containment.
If l
there are activity alarms present in the Nuclear Annex or Reactor Building, the operator should consider a LOCA outside of containment. Otherwise, the operator
[
should proceed to block #12.
Block #12 The operator is asked if the diagnosis of one event is apparent.
Since all
)
significant parameters have been considered when proceeding through the Diagnostic Flowchart, the operator should be able to answer this question.
If one event is apparent, the operator should implement the appropriate ORG. If the diagnosis of one event was not apparent, or if the operator was not sure of which event was in progress, the FRG would be implemented.
i i
For the purposes of this flowchart, an uncomplicated reactor trip is considered an event. Therefore, if the operator entered the Diagnostic Flowchart after an i
uncomplicated reactor trip, the operator would never be directed to consider any i
event.
Yet the question would be asked in block #12 if the diagnosis of one event is apparent. For an uncomplicated trip, the answer would be "YES" and the i
operator would implement the Reactor Trip Recovery (RTR) ORG.
I I
i I
i
SYSTEM 80 +"
TITLE DIAGNOSTIC ACTIONS EMERGENCY OPERATIONS Page "
of "
Revision ""
GUIDELINES Break Identification Flowchart For the purpose of the bases explanations, each question block on the Break Identification Flowchart has a unique identification letter. The letters would not be necessary for the actual plant specific E0P Break Identification Flowchart (except in the bases section if desired).
l The Break Identification Flowchart is attempting to diagnose the type (i.e., Loss Of Coolant Accident, Steam Generator Tube Rupture, Excess Steam Demand Event) and location (with respect to the containment barrier) of a suspected break, through the investigation of a series of symptoms. Therefore, it is not arranged in the order of the safety function hierarchy.
Instead, the integrated plant and integrated safety functions have been considered, and the symptoms to be investigated have been prioritized so that a determination of the break type and locations can be made in the most efficient manner.
When evaluating parameters necessary to answer the questions contained in the Break Identification Flowchart, the operator should consider past and present trends such that all available information is considered when deciding the type i
i and location of the break. For example, if a Steam Generator Tube Rupture (SGTR) i event occurs, it is possible that the main steam line radiation monitors will alarm prior to the reactor trip due to the N-16 present in the primary system a due to the significantly o er p oduct on of N-16.
In this ca e operator should still consider the fact that the main steam line radiation
[
monitors had alarmed prior to the trip (even though they are not currently in j
al arm). When this information is combined with the N-16 monitor information, the operator will answer "YES" to block (D).
That way, the operator will be appropriately directed to consider a Steam Generator Tube Rupture event.
l t
l
i
)
j SYSTEM 80+"
TITLE DIAGNOSTIC ACTIONS l
EMERGENCY OPERATIONS Page
of "
Revision "
GUIDELINES Block A The purpose of this block is to help the operator distinguish'a primary side break from an Excess Steam Demand Event (ESDE). When the plant specific E0Ps are developed, the plant has the option of either placing the question "IS SUBC00 LING' INCREASING?", or the question "IS SG PRESSURE ABNORMALLY LOW?", or both questions in this block.
If a loss of Coolant Accident (LOCA) or Steam Generator Tube Rupture (SGTR) is in progress, it is not expected that subcooling would be increasing (unless the event is a small break of a size that allows the SI system to refill the Pressurizer, thereby causing Pressurizer pressure to increase).
If the event in progress is an ESDE, subcooling should be increasing (unless the t
affected SG has already blown dry). In either case the operator should consider past and present trends to determine which event is in progress. Block (A)will l
assume that at least during the initial stages of the LOCA, or SGTR, the break is large enough to cause subcooling to be constant or to der.rease. In this case, j
the Break Identification Flowchart will direct the operator to Block (C). Block i
(A) will also assume that at least during the initial stages (blowdown phase) of
+
the ESDE, subcooling will increase and/or pressure will be abnormally low in at least the affected SG. Since these symptoms indicate an ESDE, the operator will f
proceed to Block (B).
l i
Block B This block asks if Containment pressure is increasing.
If it is, the Break Identification Flowchart identifies the event as an Excess Steam Demand Event inside of containment.
Otherwise, the ESDE is assumed to be outside of containment.
SYSTEM 80+"
TITLE otAGNOSTIC ACTIONS EMERGENCY OPERATIONS Page
of "
Revision "
GUIDELINES Block C The purpose of this block is to help the operator distinguish between a LOCA inside containment, and a LOCA outside of containment or an SGTR. If containment pressure is increasing, it is assumed that a LOCA exists inside of containment.
If containment pressure is not increasing, the operator is directed to Block (D) to determine if an SGTR exists, or if the event is a LOCA outside of containment.
^
Block D i
r The purpose of this block is to help the operator distinguish between an SGTR or a LOCA outside of containment.
If there is activity indicated in the steam-plant, a Steam Generator Tube Rupture should be considered. Alarms are not used as the sole means of determining whether or not an SGTR event is in progress because the activity levels in the secondary system may not yat be high enough to trip the alarm setpoints (except perhaps the N-16 monitors).
If there is no significant activity in the steam plant (and the activity levels are not increasing), an SGTR is not assumed to exist and it is assumed that a LOCA exists outside of containment. Otherwise, it is assumed that an SGTR is in progress.
SYSTEM 80+"
TITLE DIAGNOSTIC ACTIONS EMERGENCY OPERATIONS P a g e
of
Revision "
GUIDELINES FIGURE 3-2 OVERVIEW 0F EPG DIAGNOSTIC PROCESS STANDARD poet TMP ACmCMS penromato ir psanomu oscuo Tc Acnows 1r entuu.
no m
ovanows ovous evam messu II Jr Enono h
?
u f
DLAGMC@ S vas cowwucot no
.srse asoc 1r l
coma;s wrrn Actow or oaa ir l
- evzm, wo m
)
.