ML20062K724
| ML20062K724 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/15/1993 |
| From: | Hagan R WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-REGGD-01.101, RTR-REGGD-1.101 NA-93-0236, NA-93-236, NUDOCS 9312270064 | |
| Download: ML20062K724 (65) | |
Text
_
i-W4tLF CREEK NUCLEAR OPERATING CORPORATION December 15, 1993 Robert C. Hagan Vice President Nuclear Assurance NA 93-0236 U. S. Nuclear Regulatory C:mmission ATTN: Document Control Desk Mail Station F1-137 Washington, D.
C.
20555
Subject:
Docket No. 50-482:
Submittal of Revised Emergency Action Levels for Review and Approval Gentlemen:
This letter submits a revision to the Emergency Action Levels (EALs) for Wolf Creek Nuclear Operating Corporation (WCNOC).
These changes implement the NUMARC Guidance on EALs as endorsed by Regulatory Guide 1.101, " Emergency Planning and Preparedness for Nuclear Power Reactors,"
revision 3.
Due to the significance of changes and. based on guidance provided in the NUMARC workshops, the changes are being submitted for review and approval prior to implementation.
A review of these EAL changes has been performed in accordance with the requirements of.10 CFR'
- 50. 54 (q) and it has been determined that the changes do not decrease the effectiveness of the Wolf Creek Radiological Emergency Response Plan.
WCNOC has discussed this submittal with Mr. W. D.
Reckley, NRC Project Manager for WCNOC.
Based.on'this discussion and this submittal, WCNOC would like to implement the revised EALs by the end of the first quarter or beginning of the second quarter 1994. of this letter provides a. comparison matrix reflecting WCNOC's currently approved EALs, the NUMARC proposed EAL and the WCNOC proposed EAL.
This matrix is being supplied to aid in this review. of this letter provides a copy of procedure EPP 01-2.1,
" Emergency Classifications," which incorporates.the revised EALs.
The Radiological Emergency Response Plan will be revised within 30 days of NRC approval of the revised EALs.
Q~ 1 fl n* A <
-p13y w
PO. Box 411/ Burhngton. KS 66839 / Phone: (316) 364-8831
)
000 82
^"'*"' **"*'* "" " "
F PDR
NA 93-0236 Fage 2 of 2 If you have any questions concerning this submittal please contact me-at-(316)-364-8831, extension 4553 or Mr. Kevin J. Moles at extension 4565.
Ve q truly yours,
'l W
/
bert C. Hagan Vice President Nuclear Assurance RCH/jra Attachments (2) cc:
J. L. Milhoan (NRC), w/ attachments G. A.
Pick (hTC),. w/ attachments W. D. Reckley (NRC), w/ attachments D. B.-Spitzberg (NRC), w/ attachments L. D. Yandell (NRC), w/ attachments
1 l
I i
l l to NA 93-0237
____m
EAL MATRIX ABNORMAL RAD LEVELS NUMARC l
NEW WOLF CREEK l
OLD WOLF CREEK AU1 IER 6 (NUE)
Radioactive effluent Tech Spec release limit (s) exceeded (NUE)
AU2 RER 3 (NUE)
Rad levels or airborne which indicates a severe degradation in the control of Rad materials (e.g. increase by factor of 1000 in direct radiation reducing within facility (Alert)
NONE AAl RER 2 (Alen)
A measured or projected dose rate of one mR/Hr @ site boundary under actual meteoroligical conditions (Alen)
AA2 FH $ 2 (Alert) Fuel Bldg An irradiated fuel handling FILL 3 (Alen) CTNfr accident which breaches the cladding with release of radioactivity to containment or fuel building (Alert)
AA3 IER 7 (Alert)
Radlevels or airbome which indicates a severe degradation in the control of Rad materials (e.g. increase by factor of 1000 in direct radiation reducing within facility (Alert)
ASI RER 4 (SAE)
Measured or projected dose rate of >50mR/Hr for 1/2 hour or
>500 mR/Hr WB for 2 min. or 5 times these levels to thyroid @
site boundary under actual meteorological conditions (SAE)
AG1 RER 5 (GE)
Measured or projected dose rates of one R/Hr WB or 5 R/Hr i
thyroid at site boundary under actual meteomlogical condition I
(GE) l
?
I
EAL MATRIX FISSION PRODUCT BARRIER DEGRADATION NUMARC l
NEW WOLF CREEK l
OLD WOLF CREEK FU 1 ADM3 Same as ADM 3 F
RCS FA 1 LRCB2 See copies of old Wolf Creek FS 1 SGTF2 EAL for fission product barrier I
FG 1 SGTF 12 breach or challenge indications MSLB 3 MSLB 6 MSLB 8 FEF 2 FUEL LRCB 3 SGTF1 MSLB 1 FEF1 CONTAINMENT LRCB 4 LRCB6 LRCB 7 SGTF6 SGTF 11 SGTF 17 SG1T 19 MSLB 5 MSLB 10 FEF 3
?
FEF 4 FEF5 i
I l
i i
1 i
l
I F
-.0 INDICATIONS OFFUEL CLADDING BREACH OR CHALLENGE (PaceIof1)
Sub criticality CSFST: Red or Orange l+
F-:
U
? E.~
t OR
- 1, --
l Core Cooling CSFST: Red or Orange c
j+
ft;;
.;x.-
.[
l Heat Sink CSFST: Red
- Ii l+
- N[ '
- ct:.
t OR SAi
- sA CVCS letdown Rad monitor reading 2 High Alarm jfui; '
+
AND l E."
i:D ?. O.
Confirmed analysis indicates an INCREASE of2 63 (uCi/g) gross activity in the RCS in any 30 minute period
- n
{~_
- Rt
- EI
. Ki OR
>C1
- y-Containment Air Particulate, lodine and Radiogas monitors, Area Radiation Monitors, OR
'Ee Containment High Rad Monitors increase significantly with the RCS intact due to coolant which 1al s
r
+
L.,
i normally leaks ta containment
- ( O{
C.
l OR 5}
l H ;;
SAMPLING VERIFIES THE EXISTENCE OF, EA5:'
s l ? 600 uCi/g gross activity
- Ls l+
DL;- ~
93
- N s OR mot
- LE ' <
l ? 5 times the Tech. Spec. 3.4.8 limit of100/E bar uCi/g of cross radioactivity l4 i
^
l I ? 5 times the dose eouivalent 1 31 limits ofTech Spec Ficure 3.4.1 l+
i 1
i
?
EPP 01-2.1 Rev,12 4
Page 12 of14
+
..0 INDICATIONS OF REACTOR COOLANT SYSTEM (RCS)
BREACH OR CHALLENGE (PageIofI) l RCS Intecrity CSFST: Red or Orange j+
-C OR 8
l licat Sink CSFST: Red s
l4 I'
. s..
- -A E General:
- s.
- s; The inability to maintain normal liquid inventory within the Reactor Coolant System (RCS) by no
.U e operation of the Chemical and Volume Control System which is considered as one CCP discharg cMl the charcing header.
+
E 3 n.
. ni, Loss to Containment:
R57 Both Containment Air Particulate, Iodine or Radiogas monitors increase very rapidly.E4 to off scale high 4
C'
- U ';
E.
[
D<
- o; Loss to Steam Generator:
_ Any narrow rance steam renerator level increasine in an uncontrolled manner 4
- C' H.
L' A ;
g
< I/:-
- E?
N:
,-_ c :
' g;f
':y; I
i
't EPP 01-2.1 Rev.12 Page 13 of14 l
..0 INDICATIONS OF CONTAINMENT BREACH OR CHATLFNGE
- C l Containment Intecrity CSFST: Red or Orance
'O-l+
N I
T.
- g.
- I :-
~N-l Non-isolable leak of Main Steam Line outside Containment l+
/
f E
- N-OR P
l Loss of Containment Integrity where applicable Technical Specifications cannot be satisfied l4
- A
' S ::
.s,
- U >
- M E
1 RCS leak to interfacing system outside containment (e.g. RHR) with indication of release to p.
t emironment as indicated by:
B-
- 1) Unit Vent Rad monitor OR
.R
- 2) Radwaste Vent Rad monitor OR 4
- 3) Radiation Surveys
.A i
c
.. H.
- E' l
D.
F O
R(
C.
11 4
- - A :
L
-L' E.
N:
- G -s
[
< E.:
'D" EPP 01-2.1 Rev.12 Page 14 of14 4
h
EAL MATRIX HAZARDS AND OTHER CONDITIONS NUMARC l
NEW WOLF CREEK l
OLD WOLF CREEK j
HU1 NP 1 (NUE)
Eanhquake effects have been Eardiquake seen, heard or felt inside the PAB (NUE)
NP 6 (NUE)
Tornado touching down in PAB Tomado or Switchyard (NUE)
OH 3 (NUE)
Aircrafi crash inside PAB or OH 7 (NUE)
(Vehicle Cras10 OH 5 (NUE)
Explosion in a vital area (Explosion) resulting in major equipment damage (NUE)
OH 6 (NUE)
NONE (Turbine failure)
HU2 FR 1 (NUE)
Fire in PAB requiring offsite assistance (NUE)
HU3 OH I (NUE)
Toxic or flammable gas release width presents a danger to personnelin the PAB (NUE)
HU4 LPC/ SCI (NUE)
Confirmed security threat or attempted sabotage (NUE) liU $
OH 9 (NUE)
Judgemental authority of the i
DED/DEM HA1 NP 2 (Alert)
OBE limits exceeded: Alarm Eanhquake 98C or 98D and carthquake effects have been seen, heard or felt (Alen)
NP 7 (Alen)
Continuous winds of > 95 mph NP 5 (Alert)
(Alert). Andnatural.
Tornado and highwinds phenomenon widch threaten one fission product barrier. (Alen)
OH 4 (Alert)
Aircraft crash, missiles, fire or i
explosions causing severe i
damage to both trains of safe shutdown equipment or causes challenges to two Fission Pmduct Barriers (SAE) h i
i P
+
.1 i
l EAL MATRIX HAZARDS AND OTHER CONDITIONS NUMARC l
NEW WOLF CREEK l
OLD WOLF CREEK HA2 FR 2 (Alen)
Fire which threatens one Fission Product Barrier (Alen)
HA3 OH 2 (Alen)
Entry of uncontrollable flammable or toxic gases into a vital area (SAE)
HA4 LPOSC 2 (Alert)
Security compromise which threatens one Fission Product Banier (Alert)
HA5 LPC/SC 5 (Alert)
Evacuation of Control Room anticipated or required with control of shutdown system established from Aux.
Shutdown Panel (Alert)
HA6 OH 10 (Alert)
Judgemental Authority of the DED/DEM (Alf:rt)
HS1 LPOSC 3 (SAE)
Security compromise which threatens two Fission Product Barriers (SAE)
HS2 LPOSC 6 (SAE)
Evacuation of Control Room AND control of shutdown systems not established from t
Aux. Shutdown panelin 15 minutes (SAE)
HS3 OH I1 (SAE)
Judgemental authority of the DED/DEM (SAE)
HG1 LPC/SC 4 (GE)
Physical control of the plant is lost (GF) -
HG2 OH 12 (GE)
Judgemental authority of the DED/DEM (GE) i
-t t
l 1
i
j i
i EAL MATRIX i
i SYSTEM MALFUNCTION NUMARC l
NEW WOLF CREEK l
OLD WOLF CREEK SU 1 LEP/AC 1 (NUE)
SU 2 ADM 2 (NUE)
Any plant shutdown initiated by Tech Specs (NUE)
SU 3 LEP/AC 6 (NUE)
Unplanned loss of PK02 or more than 75% of Main Control Board annunciators for more than ;5 minutes without a major plant transient in progress.
(Alen)
NONE SU 5 LRCB 1 (NUE)
NONE SU 6 LEP/AC 8 (NUE)
Complete loss of all telephones AND the ENS (NUE)
SU 7 LEP/AC 12 (NUE)
NONE SA 1 LEP/AC 3 (Alert)
NONE SA 2 SSFM 14 (Mert)
Red or Orange path on suberiticality (Alert)
SA 3 SSFM 8 (Alen)
Mode 5; Mode 6 <23' above the flange. Loss of both trains of RHR > 15 MIN OR Loss of both trains of RHR with RCS Temp exceeding 162 F 14 Mid loop condition with RCS not intact (Alen)
SA4 LEP/AC 7 (Alen)
Unplaned loss of PK02 or more than 75% of the Main Control Board armunciators with a major plant transient in progress (e.g. any tmasient which causes a major temperature and/or pressure change in the RCS.
(SAE)
SA 5 LEP/AC 10 (Alen)
Loss of offsite power AND loss of allonsite ACpower(more than 15 min.) (SAE)
Red or Orange path on suberiticality; (used with other loss of baniers for classification)
u 1
EAL MATRIX i
1 SYSTEM MALFUNCTION NUMARC l
NEW WOLF CREEK l
OLD WOLF CREEK SS 3 LEP/AC 5 (SAE)
Loss of NK01 and NK04 (more than 15 min.)(SAE)
Mode 4: Loss both tmins of RHR and all SG levels <4%
narrow range and total available feedwater flow <260,000 lb/hr and unable to dump steam to condenser or release steam with ARVs (SAE)
Loss of both trains of RHR with the core reaching saturation condition wth containment closure not set (SAE)
OR Core uncovery with RCS not intact but containment closure maintained (SAE)
NONE SG I LEP/AC 4 (GEN)
NONE SG2 SSFM 6 (GEN)
NONE s
e EAL MATRIX.
OTHER WOLF CIEEK EALs NUS1 ARC l
NEW WOLF CREEK l
OLD WOLF CREEK NONE NONE Loss of offsite Power and loss of all onsite AC power (less than 15 min. (Alert)
NONE NONE Loss of NK01 and NK04 DC Power (less than 15 min.)
(Alert)
NONE NP3 Safe Shutdown Earthquake limits exceeded; Annunciator Windows 98A "R SPECTRUM SSE EXCEEDED" or 988 SSE in Alarm and Earthquake effects have been see, heard or felt (SAE) l' NONE NONE Natural phenomonon which threatens two Fission Product Barri<:rs (SAE)
NONE NONE Core uncovery with RCS not intact and Containment closure not set (GE)
NONE NONE Fire which threatens three Fission Product Barriers (GE)
NONE NONE Security compromise which threatens three Fission Product Barriers (GE)
NONE NONE Natural phenomenon which threatens three Fission Product Barriers (GE)
Containment LRCB 7 Containment Hi Rad Barrier SGTF 19 Monitor confirmed reading MSLB 13 10,000 R/HR (GE)
FEF5 w
- to NA 93-0237 a
3 RADIOLOGICAL EMERGENCY RESPONSE PLAN IMPLEMENTING PROCEDURE WOLF CREEK GENERATING STATION DRAFT EMERGENCY CLASSIFICATION EPP 01-2.1 NUMARC Revicion EMERGENCY PLANNING REVIEW DATE PSRC APPROVAL RECOMMENDATION DATE PRESIDENT & CHIEF EXECUTIVE OFFICER APPROVAL DATE RELEASED DATE Page 1 of 52
P 1
t
~
RADIOLOGICAL EMERGENCY RESPONSE PLAN IMPLEMENTING PROCEDURE 1.0 PURPOSE l
This procedure provides guidance to evaluate plant conditions during an actual or potential emergency situation, assess the Emergency. Action Level (EAL) exceeded and classify the emergency according to its severity.
2.0 APPLICABILITY This procedure applies to the Shift Supervisor as Duty Emergency Director (SS/DED), Duty Emergency Director (DED), and Duty Emergency Manager (DEM).
This procedure shall be implemented immediately upon recognition of an emergency or off-normal condition.
3.0 DEFINITIONS 3.1 Alert Events are in process or have occurred which involve an actual cr potential substantial degradation of the level of safety of the plant.
Any releases are expected to be limited to small fractions of the Environmental Protection Agency (EPA) Protective Action Guideline exposure levels.
3.2 Critical Safety Function Status Trees (CSFST)
For purposes of emergency classification, the following conditions are conservatively defined:
3.2.1 Red Path Barriers will be considered breached when the CSFST associated with the barrier is proceeding along a red path.
Page 2 0f 52
i RADIOLOGICAL EMERGENCY RESPONSE PLAN IMPLEMENTING PROCEDURE 3.2.2 Qrance Path Barriers will be considered to be subject to a severe challenge and shall be considered breached when listed in an individual initiating condition.
Other situations will be handled at the discretion of the DED/DEM for classification purposes.
3.3 Emeroency Action Levels (EALs)
Plant or radiological parameters which are the basis for quantifying the initiating condition and classifying the severity of the emergency.
3.4 Emercency Classification A system used to define the severity of emergencies into one of four categories based upon projected or confirmed initiating conditions / emergency action levels.
Classifications listed in order of increasing severity are:
Notification of Unusual Event, Alert, Site Area Emergency and General Emergency.
3.5 Emeroency Conditions Situations occurring which cause or may threaten to cause radiological hazards affecting the health and safety of employees or the public, or which may result in damage to property.
3.6 Exclusion Area That area surrounding the Containment Building to a distance of 1200 meters.
3.7 General Emercency Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with the potential for loss of containment integrity or the potential loss of reactor coolant system integrity.
Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Page 3 of 52
J RADIOLOGICAL EMERGENCY RESPONSE PLAN IMPLEMENTING PROCEDURE 3.8 Imminent The fact that an event will or may occur very soon or that an unavoidable event will happen even if it is in the future.
3.9 Notification of Unusual Event Unusual events are in process or have occurred which indicate a potential degradation of the level of safety of the plant.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
3.10 Protected Area That area around the plant which is encompassed by physical barriers and to which access is controlled for security purposes.
3.11 Site Area Emercency Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels except near the site boundary, i
I 1
^
l EPP 01-2.1 NUhb4RC Rev.
Page 4 of 52
o RADIOLOGICAL EMERGENCY RESPONSE PLAN IMPLEMENTING PROCEDURE 3.12 Vital Area Area of the site which contains structures required for safe shutdown of the plant.
These structures are:
Reactor Building Control Building Fuel Building Diesel Generator Building Diesel FOST Access Vaults Turbine Building (for structural framing integrity only)
Communications Corridor (for structural framing integrity only)
ESW Pump House ESW Electrical Manholes ESW Valve House ESW Access Vaults 4.0 INSTRUCTIONS 4.1 Precautiong 4.1.1,
" Initiating Conditions for Emergency Classification", cites specific conditions that denote, beneath thirteen emergency event categories, whether the emergency is to.be classified as a Notification of Unusual Event Alert, Site Area Emergency or General Emergency.
i 4.1.2 In all cases the decision to declare, upgrade, or proceed to recovery or closecut of an emergency rests with the DED/DEM.
The flowcharts are provided as guidance to assist the DED/DEM in making that decision.
In many cases a very general statement has been used in a block of the flowchart.
This was done intentionally to allow the DED/DEM flexibility to assess any undefinable parameters which may exist at the time.
1 4
Page 5 of 52 i
7 RADIOLOGICAL EMERGENCY RESPONSE PLAN lMPLEMENTING PROCEDURE i
1 4.1.3 Plant-specific operator actions required to mitigate the emergency condition are prescribed in the appropriate Emergency Procedures (EMG) or Off-Normal Procedures (OFN) and are independent of any actions required by this procedure.
4.1.4 The DED/DEM should consider the effect that combinations of initiating events have upon the emergency classification level.
That is, events if taken individually would constitute a lower emergency classification level.
However.,
collectively they may warrant a higher emergency classification level.
4.2 Use of Attachmenta 4.2.1 Start in the upper left corner of the chart to be used.
CAUTION:
Many charts have blocks that contain multiple initiating conditions separated by "gr and blocks that a
combine initiating conditions into two distinct sets "pr" plus "pr"
- "and".
l 4.2.2 Follow the arrows hnrizontally for yes statements and vertically for no statements.
4.2.3 For purposes of these flowcharts " site" is considered the Exclusion Area Boundary, " plant" is considered the Protected Area.
4.2.4 The designator at the upper right hand corner of the boxes is the reference to the bases document in.,
" Explanations / Bases for EALs", gives the reasoning for the box and should be referenced if any clarification is needed.
Page 6 of 52 O
l
~
RADIOLOGICAL EMERGENCY RESPONSE PLAN IMPLEMENTING PROCEDURE 4.2.5 Full size (approximate 11 x 14 inch), copies of Attachments 1, 2,
and 3 are maintained in the Control Room, the Technical Support Center, the Emergency Operations Facility, and the Simulator. color coding (similar to that used in.
the EMG's) is as follows:
No Action In This Category - GREEN Notification of Unusual Event - BLUE t
Alert - YELLOW Site Area Emergency - ORANGE General Emergency - RED 4.2.6 contained in the procedure is not color coded.
4.3 Initial Actions CAUTION:
Outage / shutdown conditions should be given special consideration as they are likely to create abnormalities such as the loss of RCS pressure boundary (refueling, mid-loop operations, equipment hatch open, etc.).
This type of boundary violation combined with a plant transient (loss of AC power, etc.) may create a worse situation than would be expected if the Unit was in-power operations.
l 4.3.1 Upon recognition that an abnormal or emergency condition exists, the on-duty' Shift Supervisor l
shall be immediately notified.
Recognition of the event can occur as a result of either Control Room personnel or other plant personnel observing the abnormal or emergency condition.
4.3.2 Control Room personnel shall continue to monitor i
the appropriate plant parameters and-instrument readings or any other symptoms which would be indicative of further systems degradation.
i Page 7 0f 52
t
~
RADIOLOGICAL EMERGENCY RESPONSE PLAN IMPLEMENTING PROCEDURE 4.3.3 Operators shall refer to the appropriate EMGs and take any actions called for based upon the indicated symptoms.
4.3.4 The on-duty Shift Supervisor shall report to the Control Room, if possible, and evaluate the event to determine the need for classifying the emergency condition into one of the four emergency classification levels.
4.3.5 The on-duty Shift Supervisor shall refer to of this procedure to ascertain whether or not the event fits the general description for any of the initiating conditions listed.
If the event does not fit any of these general descriptions, the on-duty Shift Supervisor should evaluate the implications of the event and, if appropriate, classify the emergency condition based upon professional judgment.
If no classification is warranted, no further action is required except to continue monitoring the event.
I 4.3.6 If the on-duty Shift Supervisor determines that the event does fit one or more of the emergency classifications listed in Attachment 1, the on-duty Shift Supervisor shall assume the role of DED as prescribed in EPP 01-1.0,
" Control Room Organization" 4.3.7 The on-duty Shift Supervisor shall declare the appropriate emergency classification and implement the necessary actions using the appropriate checklist referenced in EPP 01-1.0 " Control Room Organization".
4.4 Subsecuent Actions The Shift Supervisor and DED shall continually monitor plant conditions to determine whether a change in emergency classification is warranted.
EPP 01.2.1 NUhb4RC Rev.
Page 8 of 52
f' RADIOLOGICAL EMERGENCY RESPONSE PLAN IMPLEMENTING PROCEDURE Whenever the Shift Supervisor or DED identifies a change in an original (initiating) condition, shall be referenced to determine whether to escalate the emergency classification or proceed to recovery /closecut of the event in accordance with EPP 01-12.1 " Recovery Operations".
5.0 REFERENCES
5.1 PIR 92-0604 5.2 PDR 91-0038 5.3 PIR 92-0731 5.4 Wolf Creek Generating Station Radiological Emergency Response Plan.
5.5 EPP 01-1.0,
" Control Room Organization" 5.6 EPP 01-10.1, " Protective Action Recommendations' 5.7 EPP 01-12.1, " Recovery Operations" 5.8 Emergency Procedures (EMG) 5.9 Off-Normal Procedures (OFN) 5.10
" Technical Specifications for Wolf Creek Unit 1",
Docket No. 50-482, NUREG 1136 1
5.11 Reg. Guide 1.101, " Emergency Planning and Preparedness for Nuclear Power Reactors.
5.12 WCGS Off-Site Dose Calculation Manual 5.13 WCGS Updated Safety Analysis Report 6.0 RECORDS 6.1 NONE i
7.0 ATTACHMENTS EPP 01-2.1 NUrMARC Rev.
Page 9 0f 52
I
~
RADIOLOGICAL EMERGENCY RESPONSE PLAN IMPLEMENTING PROCEDURE 7.O ATTACHMEITTS 7.1,
" Initiating Conditions for Emergency Classifications" 7.2,
" Indications of a Loss of I
Function" 7.3,
" Explanations / Bases for EALs" l
)
l l
Page 10 of 52
ATTACllMENT I IhTUATING CONDmONS FOR EMERGENCY CONDmONS (Page 1 of 13)
Radioactive Effluent Release RER1 RER2 RER4 RERS Rapidly Meq me6ngs or Unp!annel r9ase of gawoes or Measured OR peqacted dose rate at Measured OR gyopeted he rate Hi HI Alarm c" W r h % >200 W SRe Beundry > 100 mr!br WB er 500 at Sae Bossyfry > 1Rhr WB QI3
- GT RE 210 (Chann*l 213)
- Ves - >
the Hi Hi Safoowd of the rmh thyreed 5% thy mf OB r% Proc-n Rad urrem E
08 GH RE 109 (Chennet 103) 3 15 ninutes
-Y es - b GT RE 21B (Charwwt 213)
- Yes->
GT RE 210 (Channel 213)
-Yes GT RE 218 (Channet 213) 38 7E*07 uCi. sac 18 7E*06 utvsec GH RE 10 B (Channel 103)
M QB BM RE 52 (Channel 526)
GT RE 108 (Channel 103)
GH RE 100 (Channel 103)
LE RE 59 (charmet SOS) 316E *07 uCvsec 3 6E*08vCaisac 1
Hr RE 45 (Channal 456)
HB RE 18 (Chmn*I t en)
No NO HF RE 95 (Channe1956) 98 "O
As sndcated by Cfwmistry Samde enaw s
^k k! TbhNbk k..
5[*t an ig;w v.~p@g 8Enmsse it-.
No gg
- j y
RER7 Unplanned OR Control Room Area Rad unewdained Area Red Ms VaM mW
_yes_9
- ALERT.
ygg
-~
monace increases by a
= 15 mrh f actor of 1000.
(50 RE 33)
No 4
RER6 y
Urrlanned release of gasous or lhtuid iescactM'y
- 2 tunes the Hi HesetW No of tne renovnng proces, red monam 3 60 men GT RE 218 (Channel 213)
GH RE 10B (Channel 103)
BM RE $2 (Channel 526)
LE RE 59 (Channel 596) f M' %N" - k.N oe-
~ ~ - -
HF RE 45(Channel 456) yes v
HB RE 18 (Channel 186)
' /1)fJUSUAU EVEt4TF*
HF RE 95(Channef 956)
~V"'
"2~
on As Irvficated ty thermstry sample aN
_. lnusR
$oacT@dl6%. !1
==W' vio l
tATEG 7 sn 7 - -
Page 11 of 52
NITACIIMENT I INIT1ATING CONDmONS l'OR FMERGFNCY CONDITIONS (Page 2 of 13)
Loss of Reactor Coolant Boundary LRCB2 LRCB3 i RW RC5 tweak is indKated by 8E Fahd fuelis Indmated by AM of the Both trabs of CISA or (pCBj of the foileMng for-yv,mg CISB incomplete
- RCS leak > especify of one cep CORE COOLING ORANGE PATH QB discherping to normat chargkg pg Containment Red Path y"
tJn*1-nt!ned ce Pressure boundary ANQ ptr level <4%
CORE COOttNG RED PATH 9B lenhage > 10 GI'M
_yg,
QS
_ y,3 4 QB Yesg>
Cere Coolmg Red Path 08 HEA) SINK RED PATH HEAT SINK RED PATH
>15 m idweed i-akag- > 25 cru gg gg gg INTEGRITY RED PATH
$300 utugm CEI An unemiaced pg I
gg Containment pressure GT RE 59 er 60 reading GT RE 59 er 60 resdN
- *8se P30 D.
>1 AE+03RMR t2 BE +03R/HR y
}j,..._ ;,-
..a No
/
"y LRCB7
[ LRCBS GT RE 59 or 60 reading
-Yes-
$1E*04 R/HR Si fien required ANQ total g
ECCS riew < 225 gpm NO e.
=.a.,
8 M n.~ = AREAL @h N.
LRCBS J
SIT M@bhh h Both trains of C'ISA or CISB incomplete
- N vt menC3D g3 Conlahment Red Path OB Core Cooiirs Red Path >15 min Yes RB An unexplained Containment pressure
-s deorease HALERT1 No
,~. m w,w
... e
!;NOT@lC6TICt4;OF{
? uNU$UAL EVE.gy, NTE e wgn w v,w nn w m..:.:e 5 ~
CATEGDR,
D5idi mmw W: "e',-
- Note: Unable to set proper Emergency Alignment per ESF Status Panels (CISA or CISB)
NOTE This chart shall not be used if Steam Generator Tube Rupture Failure is the only event Go to ' Steam Generator Tube Failure?
Page 12 of 52
ATTACliMENT I INTTIATING CONDITIONS FOR EMERGENCY CONDITIONS (Page 3 of 13)
Steam Generator Tube Failure SGTF1 SGTF2 SGTF3 SGTF4 Faned fuel rs ind<a!ed by any of th*
SG tube rtpure >
Uneselable SG 34,,y,,,, on f%
cepseRy of one CCP
-Y es >
tau!t outsde Yes g-SG Mb SGTR
- Core Cochng Orang = Path g.charg'ng to the CTMT SGTF6 EkAh teans of CI5A or CISB 03 normal chat ging
- Core Conkrg Re 1 Path Yes' b,,.jer @
Incmielete* 03 o
03 pressu,,,, y N
Conta'nment Red Fath 08 SGTF19 i SGTF5 Core Cooling Red Path y,,
T
- Heat Set Red Path e4g CB iEMG E-3)
> t5 rmn 9B GT RE 59 or to
. > 300 pCvgm DEL
-Yes*
USMaw Censamrnent
_ l 3tE*04 RUR r, w
..y es -
08 l
SG fatA insnie CTMT p, essure dec2*as*
No
. GT RE 59 or 60 reading 22 eE *03 RHR i SGTF7 4
I SG Tut e iceksge
>S00 GPD in any one SGTF8 soTpg y o.,, ~ -.,
MDESITE%REAWik No SG or 1 GPM tota!
--Yes >
Unisotable SG faull Yes y*
outsee CTMT SGIE1.1 SG fault is on fGid M R6ENdYt?
6 Beth trains of CtSA or CiSB SG with SGTR Ww rvc.'""
=
No SGTF10 incompicte 93 Contenment k
Red Path QB G te chad fm MSL SG lault insde CTMT
.-Yes>
Break er Fuel Failure min EE Uwa Cwam No SGTF12 l
pess decrease SG tube rtmture > ew y or SGTF13 e
NO one CCP dschargog to l
rema charges bender M E N,4 s
t,,,,, og,,,,, p,,,
go
-Yes Yes pressurizer level
- 4%
(EMG E-3) lgo SGTF18 No t SGTF14 m3 m ;
- m. ge Unisolable SG fava Yes J '-
~I E
SG tube leakage
- 500 Yes W CTMT n>
-Yes GFD in any 1 SG or RGTFD
~' ~ "
pjo 1 GIM M80 Both trains of CISA or CtSB
, SGTF16 w.e gg O
Containment Red Path QE SG faut inside
_y e _,,
Core Cochng Red Path yes LERT.'
>15 min g3 Unexplained Containrvent pressure NO l
decrease No i
No I
diwwasFMMkeq NQ MUSd Y$.y.M$h
- Note: Unable to set proper Emergency Alignment per ESF Status Panels (CISA or CISB)
Page 13 of 52
ATTACllMENT I INmATING CONDmONS FOR EMERGENCY CONDmONS (Page 4 of 13)
Main Steam Line Break MSLB1 Fased fuelis indicated by ANY of the following MSLB3 MSLB12
- Core cociing oa Path g
.Cwe Coolmg RedPath SGTR on taulted SG > capacRy
#*d $
M* 8G '*d E
E outside CTMT Yes n-,
d em CCP 65tharging to_
-Yes*
.93
-Y e s _ >
Heat Sek Red Path normal charging @
PRZ ievet < 4%
gg
.*300uCVgm DEI No 9E g
No GT RE 59 or 60 Readig gyg(gj3 No 12 BE*03R/HR g
No gota i,ams of cisA or CISO reaW 4 es v]dELB))
h MSt nd becmplete.
I1E W E R E
N,o iN M P'N V W -^~~ t Man Steam Une OR QS
-Yes-
,gh,.
maw Red Pam g
Fead Water tweak in SG f ault inswie CTMT
-Yes->
gg 4
Turbme BuMmg Yes.
gg Core Cochng Red Path > 15 rrwn WRs4-sps, gege No 93 Area 5 an unexplained Containment j
Go to cfurt for SGTR pressure decrease MSLB6 No or Fud Elernent SGM on faded SG > capach Fedure l
N,o of one CCP discharging to
'~
t MSLB7
"*"" Ch"'9'"U M PRZ level < 4%
(Jnisolable SG fad Yes i
outskie cTur No
,m MSLB10 P.. O ALER.TG -
l y
3 No Both trains of c!SA er CISB MSLB8
^^
h MSLB9 SGM on faded SG > capacy g ' y9ygy QY ]g.{
p Caitainment Red Path y
P 6hging b
- YeS-l' T.9.
d normal charging M e sI.E.3pIb; w3 SG fault inside CTMT Yes - >
- Yes -
-- x e
pay w,4g Core Cooling Red Path >15 min I
9B i
An unexplained conwnment No No s,<
1 m
gEg g y pressure decrease
. ~ _.
7 No
" ~
pg j _
u
...z m m @
~ EVRM R
m-i l
h&gqfyyy l
- NOTE: Unable to get proper alignment per ESF STATUS Panels (CISA or CISB.)
\\
l i
I EPP 01-2.1 i
NUMARC Rev.
Page 14 of 52
- i..
ATTACIIMENT I INU1A11NO CONDFIlONS FOR EMERGENCY CONDn10NS (Page 5 nr 13)
Fuel Element Failure FEF1 FEF2 FEF4 F ew fuelis indented by effI of the RCS tweak is irdcated Both trams of CISA or CISB 1rromptede *
"9#
Conta Red Fath
-RCSkak> e dy e discharging 9"U
. Core Coorg ed Path Core Coocng ed Path exists
- Yes4 93
~Yes-9
-Yes
> 15 rrun
. Hest Sink Red Path SM RM PM p3 93 An aW CotaWM
. > 300 pCVgm DEI
- Hegr4y Red Path OB pressure decree =,
. GT RE 59 or 60 rea$ng
^" '" "U I' No 32 BE*03R/HR l
4 FEF5 No GTRE $9 or 60 readng j
FEF3
>1E*04R,uR Yes Doth trams of CISA or CISB mconWe
- l 98 Conternment Red Path No I
k.yg).g[idiNNAhk[ C 8
RB Cee coot +ng Red Path erists Yes g+g g
$15 mm
,7
-,y yq 90 An unexplained Cordalnment pressure decrease i
No
[ ALERT [
FEFS u
RCS Adivdy > Tech Spec 3.48" pt3
.,n.~.
e ~.e. + v..,
SJt.016 Hi H1 sta'm and idAT!OROfs 4fiOT$sO^tEVEN6
- Y'S
- 1UNu enawis shows en increase >63 uCFgm 5 i-1"*' ' l < H' W +
gross activRy in 30 min i
no
~,..n 14NONwa
,s,.
i C11Mi lipJ MfNbI$$bNk
' Unable to get proper Emergency Alignment per ESF Status panels (CISA or CISB sections)
" When DEI exceeds limit for >48 hr during one continous interval QB exceeds the limit line on Figure 3.4-1 of TS 3.4.8 QB gross activity >100/ E BAR EPP 01-2.1 NUMARC Rev.
Page 15 of 52
ATTACIIMENT I IN!11A11NG CONDil10NS FOR EMTRGENCY CONDITIONS 6 age 6 of 13)
Loss of Electrical Power / Assessment Capabiltiy LEP/AC2 LEP/AC1 i r fc LEP/AC4
'"'"8"'"**'"s' I
toss on Fre, cred t" Se'* ],5g m
__ y es _,
_y,,_,
g,,%,,
_,,, _,j us I c ves or sac r e >i l
EMG C-0 A
4 4
4 _-
S, RO?A!TEAFfAt,e-l LEP/AC10 t
XEMERGENCW!?
t 3x;5 e 7.me; - ~
Ordy one NB Buss energ> sed
'dLERT.
e No J[q,NOTIE! CATION OFA c.,
t y!UNU$UAU EVENTO LEP!ACS LEP/AC9 Less of a l Class (E
. ;. 3..
..n., f u,
r DC m k4
-Yes - >
Modes 1 - 4
---Yes h:,EMERGENCW newsrsies we
- r-for > 15 n*i
- - - i gy +n_w-= n l
I No No
+ LEP/AC6 LEP/AC7 Unplanned bss of
- 75%
LEP/AC11 or MCs annunciaiss c.
A map trarment in bdicate I 15 muutes reogress causw a rnayw
_y Q6 IPM1p DN10F PT*sS ChSOQe Ung4asv*d loss of FK01 or in the RCS WH 00 02M PK023 15 minutes I
, LEP/AC12 No I
Unptanned ios, or 98 ALERT eperable Cle*s 1E No DC power No
_ yg3 _
NK01/3 OR NK02/4 s105 vde for > 1$ men
>e
-=>m.*
- ~--
9 LEP/AC8 1N"OTl"EICAT'I"O'N'OFe r
- IIUNOSUAl2EVENTI Compkte loss of s!! Onsde
- s ;" c - ry -, '
r E
Yes Offs #e Commuencaten capitdidy No i
No s.w, c m,
0 to AcnoNmism s
GI $d5TEddhVE.~C' h N:?:4}:1. &::i, T;~ :::
-?
Page 16 of 52
. - o ATTACitMENT 1 INIT1ATING CONDITIONS FOR EMERGENCY CONDrilONS (Page 7 of 19 Fuel Handling Accident Fila 2 Radioactivty is retessed to l
FHA 1 the fuel BuMing as mar.d try Hi Ht mani, on I
Fuet herwhig accidmt glmoreer-i In accordance with Yes
-GG RE 27 (Chamet 273)
-~Yes
-u
..... Q.3 No
-SD RE 37 (SFP Brktge
'J
^"i'ALERTl@Ov: fA OFN KE418
-GG RE 28 (Channel 263) g
-SD RE 30 (SrP area)
)
,4:p
" fr-f crane area) i No h
FHA 3 Radioactivfy is teleased to tw Renctor BuMeg as Indicated by Hi Hi Alarm on Mirnor*a
-GT RE 31 (Channel 313)
-Ci RE 32 (Channe! 323)
_gg
-SD RE 4i (Manipula'or Crane MonRor)
-50 RE 42 (containment area)
-SD RE 40 (Access Hatch area)
No jN6YI5iOIIj640d$1
' kunusuAi:EVERTF mxt-wwenyA t
Page 17 of 52
ATTACIIMENT I INrrlA11NG CONDIT10NS FOR EMERGENCY CONDT110NS (Page Sof 13)
Safety System Failure or Malfunction SSFM1 53pg2 SSFM3 Vnahte to fe=d as SG enth A8 SG NR l*Ve* '5%
+
Mav) Feed Warw (AE)
AND g3%,g tdat AFW f' w
- 260k Ibmi Yes CR w
y br totai E S flow
_ y,s.
C@ ate (AO) a pg en I
(EP.*G FR H 1)
Act F*ed WetortAL)
{
No tb
\\
SSFM6 SSFM5
,g%g y, t
SSFM4 Rex
- trip brede's Rext r Tre wwe w ennnet b. md w a the a
Sh"S fMad to try N Cordred Recen trip sw4ches
_y,,
- res -
88 g" "#3 core coeftng Red Path No SB HS42 en RLOO6 E
I tw SSFM7 y
No tevasessor nyfuncem e
needad to reach AE t_m((;j gjgggy g[3 mawain het shu dewn in
-ves r
4
)
g "c=*"c' * *'
$W~ w~s#smW MxW 2 i
SSFM9 Mo SSFMS
+
%,, % g Enter CFN EJ-015
"' N MS AND W
RCS teh ewe **ds
-~ Y es -6 E-
-Yes -
c.... s'lis
)
LALERT!(-
n e ewied to e.eeed W
yy euC*edad kt accordance with OFN EJ 15
.o C 1'Mtm21-i.k *J:
fN'O'N %' ""i"T' 0t6THla ji g
- q 4 /.
- . -
.D;N,OAfE50h,i$D.:n...S mmwr EPP 01-2.1 NUMARC Rev.
Page 18 of 52
A1TACIIMENT I INr11AUNO CONDmONS FOR D!ERGENCY CONDITIONS (Page 9ef D)
Administrative A9M1 Eccs ectuation en now to aw eve
- Y
p,.c......,..,...w..,s
[%NOTINCATIO.0F/jp<NOGUA a
l y g.:m y; & ~
No e
i._ ADM7 Plant shutdown required by Tectmient specmeations ARQ
-Yes not achieved ete action statement hme Snu'ts NJ ADM3 Containtrwnt breach by Rself wfwte the appucable tach spec action statement gg _,
cannot be satisfed No k
- 36&kdh.
xem d
EPP 01-2.1 NUM//> C Rev.
Page 19 of 52
. =
ATTACIIMENT 1 INmATING CONDmONS FOR EMERGENCY CONDmONS (Page 10 of 13)
Loss of Plant Control / Security Compromise LPCISC1 LPC/SC2 LPC/SC3 LPC/Sy
--Yes f
Sacer#y A!*rf declared rt accordance Securay E energancy Ininnion into plad vMal wth WCGS Sa aguard Cenim r
less of physalcontrof of declarad in accordarx e
- Yes ->
area by unauthertred
~ Yes-4>
han
~Y
wth WCGS Safeguard MWs the CWS Rm OJ Am Contingency Plan shutdown Panel due to unauthorized individues g,
l LPC/SC6 occupying these ereas LPC/SC5 no te
,j, i
Cor*of row, evacuation is te<phed Cmtrol estatished at g
(CF N 00 017) yes y
the AUX SHUTDOWN
-Yes "
FANEL
' 'kbbkI'.N*
with in 15 m6n No l
$ $lisilifs.R. @ n
=
~
,. 4s 8,
WIE~MERGErdCY n emesses %
,. p' ? ) q -,
j :
KN',p ;..sv.,OTWCATl'b cri,,,
y m.
iM$.50@fNNG$t
,=
ii.pxa w..w;; :: :- ~ n :n 3,..
Lg4W ae"w
Page 20 of 52
ATTAC11 MENT 1 IN111ATING CONDITIONS FOR EMERGENCY CONDITIONS (Page 11 er 13)
Fire FRt FR2 The a3sMe tfw protected area tasfW F he In any of the foboMng
= 15 minutes
_y,,_
hik2ngs:
(OFN KC.016)
-Reactor BuPdng
-Cont rd Bufksng y
-Fuet Building No
-AuxlRary BuRdng
-oiesd Ge wr*x suaanc b.
-CNesef FOST Vaut
_y,,
_,3,,,y
,"m c,g-l ALERT 3
.lm;w Turbine Bunding i
~
{omrnunications Corektor
-ESW Pumphouse
-ESW Electrical Manholes
-ESW Vafve House
-ESW Access Vaults j
I no
.,-~.,y..o_,e>m ifAQT@ CAT (OKQQ,
+10NOSUAll8VENE--
wexrw nr:+v
% ih WE kb NM :(itfR&eM 2%O
.w mw ao EPP 01-2.1 NUMARC Rev.
Page 21 of 52 i
W ATTACIIMENT I INmA11NG CONDmONS FOR EMERGENCY CONDmONS (Page12cf13)
Natural Phenomena NP1 NP2 NP3 Earthquake fen h plant CBE Umns eweeded as bdicated MQ by BBC "R SPECTRUM OBE SSE 1.imMs eeeatkd as kdcated ty 98A h
_y,,,
R SPECTRUM SSE EXCEEDED
- QS
-Yes g 3
9&E *Se4mk: Recorder
-Yet*
EXCEEDEtr 988'SSE IN ALARM-Off b e'sm QB
.p p
96D *OBE IN ALARM-i IJo flo N%
Eadhquake has esused severe l damage to safe shutdown
- " ~
equipment b eccordance WRh Attachtrent 2
~
i t4o f
Nix Continous wkwis ofE 95
-Yes ypg l
NP7 No Iornado struung wahm the plant h NPS protected aren MQ Report of visible er other htlant Indication of
~~T>
Tornade strang damage to MJ of the 4"-
within the following:
, :cG
'c d --
Prdected Area
-Beactor Bunding
" ridbM ~'I% i 3-i
-Control Buftding pj, Fuel Buikfing
-Auxiliary BuNing
.Deise! Generator Bunding W ' e-~'" ' b'b'T"IO' -+ t W --
- h N
-Dieset FOST Access VauMs
, Y(' ? M3' M ~ ' ~WUNU3tJ 3
Turbane Building (structurat i4T;t form IntegrRy)
-Communicathm Corridor (Structural form Integrity)
-ESW Pumphouse
"*",.~U
-ESW Electrical Manholes
, j[
-ESW Va've House e ' ' ' g ~ "' , j -ESW Access Vau!!s m i No i EPP 0 -2.1 NUMARC Rev. Page 22 of 52
ATTACIIMENT 1 INI11A11NG CONDITIONS IT3R EatERGENCY CONDITIONS (Page 13 of13) Other Hazards On2 CH1 E " E** t)ncordrofted ertry of teme 08 ~ 'W'**
- **W flammable ges tr4a the piart that
-Y es 0-Yes ,g g estathsh QB maintain odd shutdmn (See Attachment 2 ) I e No N f OH3 OH4 Auceaa a ash ir4o piant struchses or systems within N orotected area 3 VisNe OB olhar b plant indh.4 tion of damage to any of the No foliewing-OH5 Reactor Bunding Emploskn h the protected Confed Bunding Fuel Building area resuting in visible ~ Audary Building E damage to pe menant structures or e7;ipment Diesel Generator Building r - -.. -.. D4sel F05T Access Vault -Yes ">O FALERT:W j No Tirbine eui: ding (Structurei rarm insegrey; ,n > 1 T6g ,3 9 gyg Communientions Corridor (Shuchwal Form hiegrity) ESW Pumphouse OR Electrical Mart. oles OR Vane T urbine ianure resutung in House OR Access VauRs casing penetration or damage -Yes > M to turbine er generator seals l I No No j OH7 l$ NOTIFICATION deb - mm w m hYes-hjktgyggp4g l Train deraament en site OH12 a mMNemF No Other cuens exist wNch in the judgmert of the + OH8 SS/DED indicate: Transportation of - Yes -
- 1. Actual or Imrninent substantial core damage
-Yes OH1J OH11 with potentia! k's, of containment contantnated injured indMduas off site -Yes + 2. Potential for uncontrolled radionuclide release aber mions edst e h m Chr condhs W j Judgment of the SS/DED bdicate N'o i N h the M -Yes > actualof pety rnojor failure of plant i No of the SS/DED Indicate fMW W for ped @ of I plant safety systems k pub 5e 4 OH9 may be degraded and increased mordoring of d, @ n y e & M.41s M.hWA[ERTM M" Other condi! Ions e=fst which -Yes > plant systems is
- ${g Q ~~ 3[$E marted In the judgment of the SSI
- DED indicate a pdential degredationof thelevJIof N'o waw ~ um w safety in the plant y t h e m=wsew&M No I _a - EPP 01-2.1 NUMARC Rev. Page 23 of 52 . ~.- -.
e ATTACHhfENT 2 INDICATION OF A LOSS OF FUNCTION (Page 1 of 1) A. Indication of a Complete Loss of Function needed to achieve OR.amisin Hot Shutdown (h1 ode 4)
- 1. ALL of the following
- a. Failure to bring the Reactor suberitical with the Control Rods fully inserted.
- b. Complete loss of all Boron Injection Flow Paths.
OR
- 2. AIA of the following
- a. All SG levels <10% wide range
- b. All Steam Dumps to the condenser (AB UV 34,41,45) will not operate.
- c. All SG ARVs will not operate (AB PIC 1 A,2A,3A,4A).
- d. Complete loss of both RHR trains. (A complete loss of ESW or CCW constitutes a complete loss of RHR)
OR
- 3. The Ultimate Heat Sink (UHS) is inoperable because of ANY of the following.
- a. UllS level <1070 fL
- b. UHS temperature >90 degrees B. Indication of a loss of any function needed to maintain Cold Shutdown (hiode 5)
- 1. Complete loss of RHR (A complete loss of ESW or CCW constitutes a complete loss of RHR).
AND
- 2. Any of the following
- a. >200 degrees on any valid incore thermocouple.
- b. Uncontrolled temperature rise with no action available that will prevent approaching 200 degrees on any valid incore thermocouple.
EPP 01-2.1 NUh1 ARC Rev. Page 24 of 52 j
i - ATTACHMENT 3 t EXPLANATIONS! BASES FOR EALS (Page 1 of 28) EXPLANATIONS / BASES CHART - RADIOACTIVE EFFLUENT RELEASE RER 1.- MODES: ALL l This bos is used to indicate unexpected vent stack release rates or releases greater than ODChi allowable values. RER 2. - MODES: ALL Valid means that a radiation monitor reading has been confirmed by the operators to be correct. 'niis event escalates from the Unusual Event by escalating the magnitude of the release by a factor of 100. Pmrating the 500 mR/ year criterion for both time (8766 hr/ year), and the 200 multiplier of the associated site boundary dose rate would be 10 mR/hr. The required release duration was reduced to 15 minutes in recognition of the increased severity. Wolf Creek has eliminated efIluent technical specifications as provided in NRC Generic Letter 8+01. the corresponding maximum limit from the site's OITsite Dose Calculation hianual multiplied by 200, was used as the numeric basis of this EAL. N1onitor indications should be calculated on the basis of the methodology of the site Offsite Dose Calculation hianual (ODCht). or other site procedures that are used to demonstrate compliance with 10 CFR 20 and'or 10 CFR 50 Appendix 1 requirements - adjusted upwards by a factor of 200. Annual average meteorology should be used where allowed. RER3.- MODES: ALL Addresses unplanned increases in in-plant radiation levels that represent a degradation in the control of radioactive material, and represent a potential l degradation in the level of safety of the plant. t I l EPP 01-2.1 l NUALARC Rev. lH Page 25 of 52 l i ~,.
ATTACR\\fENT 3 EXPLANATIONS / BASES FOR EALS (Page 2 of 28) EXPLANATIONS! BASES CHART - RADIOACTIVE EFFLUENT RELEASE RER 4.. MODES: ALL Valid means that a radiation monitor reading has been confirmed by the operators to be correct. The 100 mR integrated dose in this initiating condition is based on the proposed 10 CFR 20 annual average population exposure. This value also pmvides a 4 desirable gradient (one order of magnitude) betw ccn the Alert, Site Area Emergency, and General Emergency classes. It is deemed that exposures less than this limit are not consistent with the Site Area Emergency class description. The 500 mR integrated child thyroid dose was established in consideration of the 1.5 ratio of the EPA Protective Action Guidelines for whole body and thyroid. Integrated doses are generally not monitored in real-time. In establishing the emergency action levels. a duration of one hour was assumed, and that the EAL is based on a site boundary dose of 100 mR/ hour w hole body or 500 mR/ hour child thyroid whichever is more limiting. Unit Vent and Radwaste Vent numbers were obtained using the WCGS Emergency Dose Computer Program (EDCP). [These numbers may change after Jan.1.1994 due to implementation of the new EPA 400 and 10 CFR 20 guidelines] RER 5-MODES: ALL Valid means that a radiation monitor reading has been confirmed by the operators to be correct. The 1000 m.R whole body and the 5000 mR child thyroid integrated dose are based on the EPA protective action guidance which indicates that public protective actions are indicated if the dose exceeds i Rem whole body or 5 Rem child thyroid. This is consistent uith the emergency class description for a General Emergency. This level constitutes the upper level of the desirable gradient for the Site Area Emergency. Actual meteorology is specifically identified in the initiating condition since it gives the most accurate dose assessment. Actual meteorology (including forecasts) should be used whenever possib':. i Integrated doses are generally not monitored in real-time. In establishing the emergency action levels, a duration of one hour was assumed, and that the EAL is based on site boundary doses for either w hole body or child thyroid, whichever is more limiting. Unit Vent and Radwaste Vent numbers were obtained using the WCG5 Emergency Dose Computer Program (EDCP). [These numbers may change after - Jan.1,1994 due to implementation of the new EPA 400 and 10 CFR 20 guidelines) EPP 01-2 I ' NUMARC Rev. Page 26 of 52 W e %w = +. -.
ATTACHMENT 3 EXPLANATIONS / BASES FOR EALS (Page 3 of 28) EXPLANATIONS / BASES CIIART - RADIOACTIVE EFFLUENT RELEASE RER 6. - MODES: ALL The term Unplanned", as used in this context, includes any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. Valid means that a radiation monitor reading has been confirmed by the operators to be correct. Unplann M releases in excess of two times the ODCM specification that continue for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is not the primary concern here; it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes. Therefore, it is not intended that the release be averaged over 60 minutes. For example, a release of 4 times T/S for 30 minutes does not exceed this initiating condition. Further, the Duty Emergency Director / Duty Emcrgency Manager should not wait until 60 minutes has clapsed, but should declare the event as soon as it is determined that the release duration has or willlikely exceed 60 minutes. RER 7. - MODES: ALL Valid means that a radiation monitor reading has been confirmed by the operators to be correct. This IC addresses increased radiation levels that impede necessary access to WCGS, or other areas containing equipment that must be operated manually, in order to maintain safe operation or perform a safe shutdown. It is this impaired ability to operate the plant that results in the actual or potential substantial degradation of the level of safety of the phmt. The cause and/or magnitude of the increase in radiation levels is not a concern of this IC. The Duty Emergency Director / Duty Emergency Manager must consider the source or cause of the increased radiation levels and detennine if any other IC may be involved. For example, a dose rate of 15 mR/hr in the Control Room may be a problem in itself. liowever, the increase may also be indicative of high dose rates in the Containment Building due to a LOCA. In this latter case, an SAE or GE may be indicated by the fission prod 3ct barrier matrix ICs. Areas requiring continuous occupancy includes the Control Room and the Central Alarm Station. Section Ill.D.3 of NUREG-0737, " Clarification of TMI Action Plan Requirements", provides that the 15 mR/hr value can be averaged over 30 days, the value is used here without averaging, as a 30 day duration implies an event potentially more significant than an Alert. EPP 01-2.1 NUMARC Rev. Page 27 of 52 m .mm. -m m .m mm.m-..s ,m.____ m -,.v-u. c-., -, e ,~ ed
. ATTACIBIENT 3 EXPLANATIONS / BASES FOR EALS (Page 4 of 28) EXPLANATIONS / BASES CHART - LOSS OF REACTOR COOLANT BOUNDARY i NOTE: The note is provided to direct the user to the Steam Generator Tube Failure chart vice Loss of Reactor Coolant Boundary when Reactor Coolant leakage is via the Sicam Generator tubes only. LRCB 1.- MODES: 1 TilROUGli 4 This IC is included as an Unusual Event because it may be a precursor of more serious conditions and, as result, is considered to be a potential degrndation of the level of safety of the plant. The 10 gpm value for the miidentified and pressure boundary leakage was selected as it is observable with normal Control Room indications. Lesser values must generally be determined through time-consuming surveillance tests. This EAL for identified leakage in comparison to unidentified or pressure botmdary leakage. LRCB 2.- MODES: 1 TIIROUGli 4 1. Critical Safety Function Status: This EAL is for using Critical Safety Function Status Tree (CSFST) monitoring and functional recovery procedurcs. RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings, and these CSFs indicate a potential loss of RCS barrier
- 2. RCS Leak Rate: De " Potential Loss" EAL is based on the inability to maintain normal liquid inventory within the Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System which is considered as one cenuifugal charging pump discharging to the charging header. Pressurizer level was added because this is Safety hijection Criteria per WCGS procedure OFN BB-007 "SG/RCS Leakage fligh".
- 3. Containment Radiation Monitoring: The 1.4 E +3 R/hr, reading is a value which indicates the release of reactor coolant to the containment. De reading assimics the instantaneous release and dispersal of the reactor coolant concentrations (i.e., within T/S) into the containment atmosphere. Per Tech Spec 3.4.8 we can operate up to 48 hours with DEI of 63 pc/gm. His is equal to 1% failed fuel per WCGS USAR. De readings were obtained from WCGS EPP 01-2.4 " Core Damage Assessment Methodology" Attachment 1.0 for failed fuel.
EPP 01-2.1 NUMARC Rev. Page 28 of 52 %. m s. .mi.. u m .2 . m - m m..
ATFACHMENT 3 EXPLANATIONS / BASES FOR EALS (Page 5 of 28) EXPLANATIONS / BASES CHART - LOSS OF REACTOR COOLANT BOUNDARY LRCB 3.- MODES: 1 TIIROUGli 4 1. Critical Safety Function Status : This EAL is for using Critical Safety Function Status Tree (CSFST) monitoring and functional recovery proced', RED path indicates an extreme challenge to the safety function. ORANGE path indicates a severe challenge to the safety function. Core Cooling - ORANGE indicates subcooling has been lost and that some clad damage may occur. Heat Sink - RED indicates the ultimate heat sink function is under extreme challenge and thus these tu o items indicate potential loss of the Fuel Clad Barrier. Core Cooling - RED indicates significant superheating and core uncovery and is considered to indicate loss of the Fuel Clad Barrier.
- 2. Primary Coolant Activity Lercl: The 300 Ci/cc DEI assessment by the NUh1 ARC EAL Task Force indicates that this amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2% to 5% fuel clad daInage This amount of clad damage indicates significant clad heating and thus the Fuel Clad Barrier is considered lost.
- 3. Containment Radiation hionitoring: This reading is a value w hich indicates the release of reactor coolant. with elevated actisity indicative of fuel damage. into the containment. The reading should be calculated assuming the instantaneous release and dispersal of the reactor coolant nobic gas and iodine inventory associated with a concentration of 300 Ci/gpm dose equivalent I-131 into the containment atmosphre. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allow ed within technical specifications and are therefore indicative of fuel damage (approximately 2-5% clad failure depending on core inventory and RCS volumet To be conservative. 2 % clad failure and'10 hours after shutdown were selected from WCGS EPP 01-2.4. " Core Damage Assessment hfethodology",.0.
4 EPP 01-2.1 i NUMARC Rev. Page 29 of 52 l
ATTACHMENT 3 EXPLANATIONS / BASES FOR EALS (Page 6 of 28) EXPLANATIONS / BASES CHART - LOSS OF REACTOR COOLANT BOUNDARY LRCB 4 - MODES: 1 TIIROUGli 4 Containment Isolation Valve Status After Contaimnent Isolation : His EAL is for using Critical Function Status Tree (CSFST) monitoring and functional recovery procedures. RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings and/or sampling results. and thus represents a potential loss of containment. Conditions leading to a containment RED path result from RCS barrict and ior Fuel Clad Barrier Loss. Thus, this EAL is primarily a discriminator between Site Area Emcrgency and General Emergency representing a potential loss of the third barrier. In this EAL,. the function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety functions. He procedure is considered efTective if the temperature is decreasing or if the vessel level is increasing. The conditions in this potential loss EAL represent imminent melt sequence w hich, if not corrected, could lead to vessel failure and an increased potential for containment failure. In conjunction with the core exit thermocouple EALs in the Fuel and RCS barrier colunms, this EAL would result in we declaration of a General Emergency -loss of two barricts and the potential loss of a third. If the function restoration procedures are ineffective, there is no " success" path. Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation within the reactor vessel in a significant fraction of the core damage scenarios. and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow function restoration procedures to arrest the core melt sequence. Whether or not the procedures will be effective should be apparent within 15 minutes. nc Duty Emergency Director / Duty Emergency Manager should make the declaration as soon as it is determined that the procedures have been, or will be ineffective. He reactor vessel level chosen should be consistent with the cmergency response guides applicable to the facility. LRCB 5 - MODES: 1 TIIROUGil 4 His IC used to determine if any ECCS System is capable of delivering sufTicient volume of water to the core. 225 gpm was chosen because it is conservatively larger then Tech Spec delta P requirement of =210 gpm at 2400 PSID. l LRCB 6 - MODES: 1 TIIROUGII 4 See LRCB 4. l LRCB 7 - MODES: 1 TIIROUGli 4 This reading is a value which indicates significant fuel damage w cll in excess of the EALs associated with both loss of Fuct Clad and loss of RCS Barriers. As stated in Section 3.8 of Reference 5.11, a major release of radioactivity requiring offsite protective actic from core damage is not possible unicss a major failure of fuel cladding allows mdioactive material to be released from the core into the reactor coolant. Regardless of whether Containment is challenged, this amount of activity in Containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of Containment, such that a General Emergency declaration is wammted. NUREG-1228. " Source Estimations During Incident Response to Severc Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20% EPP 01-2.1 NUMARC Rev. Page 30 of 52 m m ~ m. -e--. ~. ~
ATTACHMENT 3 EXPLANATIONS / BASES FOR EALS (Page 7 of 28) EXPLANATIONS / BASES CHART - STEAM GENERATOR TUBE RUPTURE SGTF 1 - > LODES: 1 TilROUGil 4 L nis FAL is for using Critical Safety Function Status Tree (CSFST) monitoring and functional recovery procedures. RED path indicates an extreme challenge to the safety function. ORANGE path indicates a severe challenge to the safety function. Core Cooling-ORANGE indicates subcooling has been lost and that some clad damage may occur. Heat Sink-RED indicates the ultimate heat sink function is under extreme challenge and thus these two items indicate potential loss uf the Fuel Clad Banier. Core Cooling-RED indicates significant superheating and core uncoverv and is considcred to indicate loss of the Fuci Clad Barrier.
- 2. Primary Coolant Activity Level: The 300 Ci/cc DEI assessment by the NUMARK EAL Task Force indicates that this amcunt of coolant activity is wcil above that espected for iodinc spikes and corresponds to about 2% to 5% fuel clad damage. This amount of clad damage indicates significant clad heating '
and thus the Fuel Clad Barrier is considered lost.
- 3. Containment Radiation Monitoring: This reading is a value which indicates the release of reactor coolant, with clevated activity indicative of fuel damage.
into the containment. He reading should be calculated assuming the instantaneous release and digesat of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 Cligpm dose equivalent 1-131 into the containment atmosphem. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel danuge (approximately 2-5% clad failure depending on core inventory and RCS volumek IGTF 2.- 310 DES: 1 TilROUGli 4 RCS Leak Rate EAL is based on the inability to maintain normal liquid inventorv withm the Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System which is considered as one centrifugal charging pump discharging to the charging header. Presstnizer level was added because this is Safety injection Criteria per WCGS procedure OFN BB-007 "SG/RCS Leakage Hich" SGTF 3. - MODES: 1 THROUGH 4 A check for S/G fault is made to determine if the next fission product boundary is under challenge or lost. The release path looked far is either a faulted. ruptured SJG or a faulted S/G to a challenged Containment. Unisolable means that the steam release from the faulted S/G cannot be stopped until the S/G has blown dn. SGTF 4. - MODES: 1 THROUGH 4 Once a faulted S/G has been determined, a release path via a faulted, ruptured S/G is checked. l SGTF 5. - MODES: 1 THROUGH 4 nis box checks for an unisolable secondary side steam release to the Containment atmosphere. l SGTF6.-MODES:1 THROUGH 4 SeeLRCB 4 l EPP 01-2.1 NUMARC Rev. Page 31 of 52
- k. -
u -m,. e .n n. s -n
m ATTACIB1ENT 3 EXPLANATIONS / BASES FOR EALS (Page 8 of 28) EXPLANATIONS / BASES CHART - STEAM GENERATOR TUBE RUPTURE SGTF7. - MODES: 1 THROUGH 4 SG Tube leakage in excess of Tech Spec limits (Tech Spec 3.4.6.2) SGTF 8. - MODES: 1 THROUGH 4 See SG1F 3 l SGTF9. - MODES: 1 THROUGH 4 See SGTF 4 j. SGTF 10. - M_ ODES: 1 THROUGH 4 See SGTF 5 l SGTF 11. - MODES: 1 THROUGH 4 See LRCB 4 l %GTF 12. - MODES: 1 THROUGH 4 See SGTF 2 l SGTF 13. - MODES: 1 THROUGH 4. With a S!G tube rupture of several hundred gpm in progress and no ofTsite power available, a singic failure of a DG would present severe challenges to S/G and fuel integrity. Limited RCS make-up capability and difriculty of RCS pressure and temperature control could lead to core uncovery and/or S/G overtill. EPP 01-2.1 NUMARC Rev. Page 32 of 52 .m m. w. s. ~- m
ATTACHMENT 3 EXPLANATIONS / BASES FOR EALS (Page 9 of 28) EXPLANATIONS / BASES CHART - STEAM GENERATOR TUBE RUPTURE SGTF 14. - MODES: 1 -THROUGH 4 See SGTF 3 l> SGTF 15. - MODES: 1 THROUGH 4 See SGTF 7 l lSGTF 16. - MODES: 1 THROUGH 4 See SGTF 5 l i . lSGTF 17. - MODES: 1 THROUGH 4 See LRCB 4 l ! SGTF 18. - MODES: 1 THROUGH 4 See SGTF 4 l SGTF 19. - MODES: 1 THROUGH 4 See LRCB 7 l EPP 01-2.1 -i NUMARC Rev. Page 33 of.52 4 .m.-. ________________________m_____4,_ _.y_. ,e, s._ .._.,y.,
ATTACHMENT 3 EXPLANATIONS / BASES FOR EALS (Page 10 of 28) EXPLANATIONS / BASES CHART - MAIN STEAM LINE BREAK MSLB 1. - MODES: 1 THROUGH 4 1. Critical Safety Function Status: This EAL is for using Critical Safety Fimction Status Tree (CSFST) monitoring and fimetional recovery procedures. RED path indicates an citreme challenge to the safety function. ORANGE pau indicates a severe challenge to tue safety function. Core Cooling - ORANGE indicates subcooling has been lost and that some clad damage may occur. Heat Sink - RED indicates the ultimate heat sink function is under extreme challenge and tims these two items indicate potential loss of the Fuel Clad Barrier. Core Cooling - RED indicates significant superheating and core uncovery and is considered to indicate loss of the Fuel Clad Barrier.
- 2. Primary Coolant Aaivity Level: The 300 uCi/cc DEI assessment by the NUMARC EAL Task Force indicates that this amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2% to 5% fuct clad damage. This amount of clad damage indicates significant clad heating i
and thus the Fuel Clad Barrier is considered lost.
- 3. Containment Radiation Monitoring: This reading is a value which indicates the release of reactor coolant. with cicvated activity indicative of fuel damage.
into the Containment. The reading should be calculated assuming the instantaneous release and di9ersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 uCilgm dose equivalent 1-131 into the Containment atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage (approximately 2 - 5% clad failure depending on core inventory and RCS volume). MSLB 2. - MODES: 1 THROUGH 4 A check for S/G fault is made to determine if the next fission product boundary is under challenge or lost. The release path looked for is either a faulted, ruptured S/G or a faulted S/G to a challenged Containment. Unisolable means that the steam release from the faulted S/G cannot be stopped until the S/G has blown dry. 5 MSLB 3. - MODES: 1 THROUGH 4 Once a faulted S/G has been determined, a release path via a faulted, ruptured S/G is checked. Leakage greater than charging capacity of one CCP to a S/G constitutes failure of the FCS fission product boundary MSLB 4. - MODES: 1 THROUGH 4 This box checks for an unisolable secondary side steam release to the Containment atmosphere. l EPP 01-2.1 NUMARC Rev. Page 34 of 52 +. - - _ _ _ _ - -. _ _ - _ _ - - - - - - _ _ _ - _ _ _ - - _ _ _ _ _ _, _ -. - _ -... - _ _ -. _ - ~ < -.,,, s
ATTACHMENT 3 EXPLANATIONS / BASES FOR EALS (Page 11 of 28) EXPLANATIONS / BASES CHART - MAIN STEAM LINE BREAK MSLB 5. - MODES: 1 THROUGH 4 See LRCB 4 l MSLB 6. - MODES: 1 THROUGH 4 See MSLB 3 l MSLB 7. - MODES: 1 THROUGH 4 See MSLB 2 l MSLB 8. - MODES: 1 THROUGH 4 See MSLB 3 l MSLB 9. - MODES: 1 THROUGH 4 See MSLB 4 l MSLB 10. - MODES: 1 THROUGH 4 See LRCB 4 l MSLB 11. - MODES: 1 THROUGH 4 Rapid depressurization of the secondary due to a MSL break u hich is isolable from the S/G's. (i.e. downstream of the MSIV's) A main steam line or feed water break in the Turbine Building or Area 5 could cause a potential degradation of the level of safety of the plant. MSLB 12. - MODES: 1 THROUGH 4 This procedure piuvides actions to identify and isolate a faulted steam generator. A Main Steam Line break inside or outside containment could cause a potential degradation of the level of safety of the plant. MSLB 13. - MODES: 1 THROUGH 4 See LRCB 7. l 4 EPP 01-2.1 NUMARC Rev. Page 35 of 52 t. -.m ~ m. ~ -. - - ,,w-e, ,..-s
ATTACILMENT 3 EXTLANATIONS/ BASES FOR EALS (Page 12 of 28) EXPLANATIONS / BASES CHART - FUEL ELEMENT FAILURE FEF 1. - MODES: ALL 1. Critical Safety Function Status: This EAL is for using Critical Safety Function Status Tree (CSFST) monitoring and functional recovery procedures. RED path indicates an extreme challenge to the safety function. ORANGE path indicates a severe challenge to the safety fmiction. Core Cooling - ORANGE indicates subcooling has been lost and that some clad damage may occur. Heat Sink - RED indicates the ultimate heat sink function is under extreme challenge and thus these two items indicate potential loss of the Fuel Clad Barrier. Core Cooling - RED indicates significant superheating and core uncovery and is considered to indicate loss of the Fuct Clad Banier.
- 2. Primary Coolant Activity Level: The 300 uCi/cc DEI assessment by the NUMARC EAL Task Force indicates that this amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2% to.5% fuct clad damage. This amount of clad damage indicates significant clad heating and thus the Fuct Clad Barrier is considered lost.
- 3. Containment Radiation Monitoring: This reading is a value which indicates the release of reactor coolant. with clevated activity indicative of fuel damage, into the containment. The reading should be calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 uCi/gm dose equivalent I-131 into the containment atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage (approximately 2 - 5% clad failure depending on core inventory and RCS volume).
FEF 2. - MODES: ALL 1. Critical Safety Function Status: This EAL is for using Critical Safety Function Status Tree (CSFST) monitoring and functional recovery procedures. RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings, and these CSFs indicate a potential loss of RCS banier. t
- 2. RCS Leak Rate: The " Potential Loss" EAL is based on the inability to maintain normal liquid inventory within the Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System u hich is considered as one centrifugal charging pump discharging to the charging header.
Pressurizer level added because this is safety injection criteria per WCGS procedure OFN BB-007, "SG/RCS Leakage High". EPP 01-2.1 NUMARC Rev. Page 36 of 52 m .. m .m__._, w -.,_ e r.-_ -m e, ..-.r-.~ r.. ~..
ATTACIIMENT 3 EXPLANATIONS / BASES FOR EALS (Page 13 of 28) EXPLANATIONS / BASES CilART - FUEL ELEMENT FAILURE FEF 3 & 4. - MODES: ALL See LRCB 4 l FEF 5. - MODES: ALL This IC is included as an Unusual Event because it is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This EAL addresses coolant samples exceeding coolant technical specifications for iodine spike. Escalation of this IC to the Alert level is via the Fission Product Barrier Degradation Monitoring ICs. lFEF 6.-MODES: ALL See LRCB 7 l EPP 01-2.1 NUMARC Rev. Page 37 of 52
ATTACHMENT 3 1 EXPLANATIONS / BASES FOR EALS (Page 14 of 28) EXPLANATIONS / BASES CilART - LOSS OF ELECTRICAL POWER / ASSESSMENT CAPABILITY LEP/AC 1 - MODES: ALL Prolonged loss of AC power reduces required redundancy and potentially reduces the level of safety by rendering the plant more vulnerable to a complete Loss of AC Power (Station Illackout). Filleen minutes was selected as a threshold to exclude transient or momentary power losses. LEP/AC 2 - MODES: 5. 6. & E I oss of all AC power compromises all plant safety systems requiring electric power, including RIIR. ECCS. Containment lleat Removal, and the Ultimate Ileat Sink. Prolonged loss of all AC power will cause core uncovering and loss of containment integrity, thus this event can escalate to a General Emergency. The 15 minute time duration was selected to exclude transient or momentary power loss. LEP/AC 3 - MODES: 1 TilROUGli 4 Loss of all AC power comprumises all plant safety systems requiring electric power including RIIR. ECCS. Containment Ileat Removal, Spent ruel 1Icat Removal and the Ultimate lleat Sink. When in cold shutdown, refueling, or defueled mode the event can be classified as an Alert, because of the significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one of the emergency busses, relative to that specified for the Site Area rmergency EA1 Escalating to Site Area Emergency,if appropriate, is by Radioactive E0luent Release, or Duty Emergency Director / Duty Ducrgency Manager Judgment ICs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. LEP/AC 4 - MODES: 1 TIIROUGli 4 Loss of all AC power compromises all plant safety systems requiring electric pow er including RIIR, ECCS, Containment IIeat Removal and the Ultimate IIeat Sink. Prolonged loss of all AC power will lead to loss of fuel clad. RCS. and containment. The 4 hours to restore AC power was based on the site black.out coping analysis performed in conformance w ith 10 tTR 50.63 and Reg. Guide 1.155, " Station Blackout". Although this IC may be viewed as redundant to the Fission Barrier Degradation IC, its inclusion is necessan ' ' esure timely recognition and emergency response. 1his IC is specified to assure that in the unlikely event of a ;. caged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as in appropriate, based on a reasonable assessment of the event trajectory. The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions. In addition, under these conditions, tission product barrier monitoring capability may be degraded. Although it may be difficult to predict when power can be restored, it is ner.essary to give the Duty Dnergency Director / Duty Emergency Manager a reasonable idea of how quickly they may need to declare a General Emergency bar,ed on two major considerations:
- 1. Are there any present indications that core cooling is already degraded to the point that Loss or Potential Loss of Fission Product Barriers is IMMININI? (CSFST shows Red or Orange path on Core Cooling OR Red path on IIcat Sink)
- 2. If there are no present indications of such core cooling degradation, how likely is it that power can be restored in time to assure flut a loss of two barriers with a potential loss of the third barrier can be prevented ?
Thus. indication of continuing core cooling degradation must be based on Fission Product 11arrier monitoring with particular emphasis on Duty Emergency Director / Duty Emergency Manager judgment as it relates to IMMINENT Loss or Potential Loss of fission product barriers and degraded ability to monitor fission product barriers. EPP 01-2.1 NUMARC Rev. Page 38 of 52 k. '. m mm m.- r + ,--v ..,_,.m-,
ATTACHRIENT 3 EXPLANATIONS / BASES FOR EALS (Page 15 of 28) EXPLANATIONS / BASES CHART - LOSS OF ELECTRICAL POWER /ASSESShlENT CAPABILITY LEP/AC 5 - MODES: 1 TIIROUGII 4 Loss of all DC power compromises ability to monitor and control plant safety functions Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is signincant decay heat and sensible heat in the reactor system Escalation to a General Emergency would occur be Abnormal Rad Levels / Radiological EfIluent. Fission Product Barrier Degradation or Duty Emergency Director / Duty Emergency Manager Judgment ICs. Fifteen minutes was selected as a threshold to exclude trainient or momentary power losses. 105 VDC bus voltage was based on the minimum bus voltage necessary for the eperation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset ofinability to operate those loads. This voltage is usually near the minimum voltage selected when battery sizing is performed. Typically the value for the entire battery set is approximately 105 VDC. For a 60 cell string of batteries the cell voltage is 1.75 Volts per cell. LEP/AC 6 - MODES: 1 TIIROUGli 4 This IC and its associated EAL are intended to recognize the difliculty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment. Quantification of "Most" is arbitrary, how ever, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perfonn a detailed count of the instrumentation lost but use the value as a judgment threshold for determining tue seventy of the plant conditions. This judgment is supported by the speciEc opinion of the Shift Supervisor that additional eperating personnel will be required to provide increased monitoring of system operation to safely operate the unit. While f ailure of a large portion of armunciators is more likely than a failure of a hrge portion ofindications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions.1he loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specincation imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50,72, if the shutdown is not in compliance with the Technical Specification action, the Unusual Event is based on ADM 2 " Inability to Reach Required Shutdown Within Technical Specincation Limits." Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. " Unplanned" loss of anmmciators or indicator excludes scheduled maintenance and testing activities. LEP/AC 7 - MODES: 1 TIIROUGli 4 This IC and its associated EAL are intended to recognize the difficulty associated with monitoring changing plant conditions uithout the use of a major portion of the anmmciation or indication equipment during a transient. Recognition of the availability of computer based indication equipment is considered (SPDS, plant computer, etc.). "Signincant Transient
- includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thctmal power change, ECCS iniections or thermal power oscillations of 10% or greater.
EPP 01-2.1 NUMARC Rev. Page 39 of 52 __________.___._______.___.m.,.___m_ m m-
ATTACilMENT 3 EXPLANATIONS / BASES FOR EALS (Page 16 of 28) EXPLANATIONS / BASES CliART - LOSS OF ELECTRICAL POWER / ASSESSMENT CAPABILITY LEP/AC 8-DC Power-MODES: ALL Communications -The purpose of this IC and its associated EALs is to recognize a loss of communications capability that either defeats the plant operations stalT ability to perform mutine tasks necessary for plant operations or the ability to communicate problems with offsite authorities. The loss of otTsite communications abihty is expected to be significantly more comprehensive than the condition addressed by 10 CII 50.72. 1he onsite communications loss includes all of the following: L Complete failure of the plant telephone system.
- 2. Complete failure of the Gaitronics system.
- 3. Complete failure of the plant radio system.
The offsite communications loss includes all of the following:
- 1. Complete failure of the ENS line.
- 2. Complete failure of offsite telephone service (inability to receive or call a location otTsite).
- 3. Complete failure of onsite fax machines.
This EAL is intended to be used only when extraordinary means are being utilized to make communications possible ( relaying ofinformation from radio transmissions, individuals being sent to ofTsite locations. etc.) LEP/AC 9 - MODES : N/A LEP/AC 5 only applicable in Modes 1 - 4. Unplanned loss of DC Power in Modes 5 & 6 is covered in LEP/AC 8. LEP/AC 10 -MODES: I TIIROUGII 4 This IC and the associated EAL are intended to provide an escalation from LEP/AC 1. The condition indicated by this IC is the degradation of the otTsite and onsite power systerns such that any additional single failure would result in a station blackout. This condition could occur due to a loss of o!Tsite power with a concurrent failure ofone emergency generator to supply power to its emergency busses. The subsequent loss of this single power source would escalate the event to a Site Area Emergency in xcordance with LEP/AC 2. LEP/AC 11 -MODES: 1 TIIROUGli 4 'this IC and its associated EAL are intended to recognize the inability of the control room stair to monitor the plant response to a transient. A Site Area Emergency is censidered to exist if the Control Room staff cannot monitor safety functions needed for protection of the public. Indications needed to monitor safety functions necessary for protection of the public must include Control Room indications, computer generated indications and dedicated anmmciation capability. The specific indications should be those used to determine such functions as the ability to shut down the reactor, maintain the core cooled and in a i coolable geometry, to remove heat from the core, to maintair the reactor coolant system intact, and to maintain Containment intact. 1 l EPP 01-2.1 l NtBIARC Rev. l Page 40 of 52
A'ITACHMENT 3 EXPLANATIONS / BASES FOR EALS (Page 17 of 28) EXPLANATIONS / BASES CIIART - LOSS OF ELECTRICAL POWER /ASSESShlENT CAPABILITY LEP/AC 12 -310 DES: 1 T11ROUGli 4 lhe purpose of this IC and its associated EALs is to recognize a loss of DC power compromising the ability to monitor rmd control the removal of decay heat during Cold Shutdown or Refueling operations. This IML is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss. Unplanned is included in this IC and IML to preclude the declaration of an emergency as a result of planned maintenance activities. Routinely plants will perform maintenance on a Train related basis during shutdown period $. It is intended that the loss of the operating (operable) train is to Iw considered. 105 VDC bus voltage was based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value should incorporate a margin of at least 15 minutes of operation before the onset ofinability to operate those loads. This voltage is usually near the minimum voltage selected u hen battery sizing is performed. Typically the value for the entire battery set is approumately 105 VDC. For a 60 cell string of batteries the cell voltage is 1.75 Volts per cell. i EPP 01-2.1 NUMARC Rev. Page 41 of 52 ..m m m __g av .c.m-Sw , ~ .--,e,
- s. _,
y .e.m- ,i,,__. 7,y
m ATTACIIMENT 3 EXPLANATIONS / BASES FOR EALS (Page 18 of 28) EXPLANATIONS / BASES CIIART - FUEL IIANDLING ACCIDENT Fila 1. - SlODES: ALL Ris procedure provides the necessary instmetions to minimize the release of airbome activity following a fuel handling accident which indicates a potential degradation of the level of safety of the plant. FIIA 2 & 3. - 510 DES: ALL NUREG-0818,
- Emergency Action Lcrels for Light Water Reactors," forms the basis for these EALs. Here is time available to take corrective actions. and there is little potential for substantial fuel damage. In addition, NUREG/CR-4982. " Severe Accident in Spent Fuel Pools in Support of Generic Safety issue 82," July 1987, indicates that even if corrective actions are not taken, no pmmpt fatalitics are predicted and that risk of injury is low.
Thus, an Alert Classification for this event is appropriate. Escalation. if appropriate, would occur via Abnormal Rad Level / Radiological Effluent or Emergency Director Judgment. i l l l i l t l l l EPP 01-2.1 l NUMARC Rev. Page 42 of 52 -,-.,..-,--.s 3- -<yr vfry-3. rrwe d--- - +- wm +wa--- P'*6r w--'vwrd v-,w. w -w w -w-g"= r'w we-=6- -+- + -- w -me +-e +-g it~ww w * - -- e-w-
- -~re------wev-
- - -- - - - ----J
ATTACHMENT 3 EXPLANATIONS / BASES FOR EALS (Page 19 of 28) EXPLANATIONS / BASES CIIART - SAFETY SYSTEM FAILURE OR MALFUNCTION SSF31 1. - 310 DES: 1 TilROUGII 3 A complete loss of secondary heat sink is indicated. l SSF312. - SIODES: 1 TIIROUGli 3 in combination with a loss of h!FW capability (addressed in box SSFht 1) a complete loss of secondary heat sink is indicated. CSFST indicators are used as determining factors. SSF313. - 310 DES: 1 TIIROUGII 3 His box is used to determine if any ECCS system is capable of delivering nater to the core. 250 GPNIis chosen because it is greater than the flow from one CCP at the PRZR PORV lift setpoint. It is anticipated that conditions leading to this box will require operator initiation of RCS bleed and feed in accordance with Eh1G frill. FRG actions should raise ECCS injection flow to that required for adequate heat removal. SSF314 - 310 DES: 1 This condition indicates faihire of the automatic protection system to scram the reactor. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient and thus the plant safety has been compromised, and design limits of the fuct may have been exceeded. An Alcit is indicated because conditions exist that lead to potential loss of fuct clad or RCS. Reactor protection system setpoint being exceeded (rather than limiting safety system setpoint being exceeded) is specified here because failure of the automatic protection system is the issue. A manual scram is any set of actions by the reactor operator (s) at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor suberitical (e.g., reactor trip switch). Failure of manual scram would escalate the event to a Site Arca Emergency. e SSF315.- SiODES: 1 Automatic and manual scram are not considered successful if action away from the reactor control console was required to scram the reactor. Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Arca Emergency is indicated because conditions exist that lead to imminent loss or potential loss of both fuel clad and RCS. Although this IC may be viewed as redundant to the Fission Pmduct Banier Degradation IC. its inclusion is necessary to better assure timely recognition and emergency response. Escalation of this event to a General Emergency would be via Fission Product Banier Degradation or Duty Emergency Director / Duty Emergency hianager Judgment ICs. EPP 01-2.1 NUMARC Rev. i' age 43 of 52 ..-.-c. m. .-.._--.--_~,r.-' s.~. .~..~.. .. m m. .m... m, m m
ATTACHA1ENT 3 EXPLANATIONS / BASES FOR EALS (Page 20 of 28) EXPLANATIONS / BASES CHART - SAFETY SYSTEh1 FAILURE OR hlALFUNCTION SSFM 6. - SIODES: 1 Automatic and manual scram are not considered successful if action away from the reactor control console is required to scram the reactor. Under the conditions of this IC and its associated EALs, the efforts to bring the reactor suberitical have been unsuccessful and, as a result, the reactor is producing more heat than the maximum decay heat load for u hich the safety systems were designed. Although there are capabilitics away from the reactor contrut console, such as emergency boration, the continuing temperature rise indicates that these capabilitics are not effective. His situation could be a precursor for a core melt sequence. De extreme challenge to the ability to cool the 're is intended to mean that the core exit temperatures are at or approaching 1200eF or that the reacta vessel water level is below the top of active fuel. For L ~ STs, this EAL equates to a Core Cooling RED condition. Another consideration is the inability to initially remove heat during the early stages of this sequence. If emergency feedwater flow is insufRcient to remove the amount of heat required by design from a least one steam generator, an extreme challenge should be considered to exist. His EAL equates to a licat Sink RED condition on the CSFSTs. In the event either of these challenges exist at a time that the reactor has not been brought below the power associated with the safety system design (typically 3 to 5% power) a core melt sequence exists. In this situation, core degradation can occur rapidly for this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier matrix declaration to permit maximum offsite intervention time. SSFM 7. - MODES: 1 TilROUGil 4 This EAL addresses complete loss of ftmetions, including ultimate heat sink and reactivity control, required for hot shutdown with the reactor at pressure and temperature. Under these conditions. there is an actual major failure of a system intended for protection of the public. Hus. declaration of a Site Area Emergency is warranted. Escalation to a General Emergency would be via Abnormal Rad Levels / Radiological Efnuent, Duty Emergency Director / Duty Eme gency Manager Judgment, or Fission Product Barrier Degradation ICs. EPP 01-2.1 NUh1 ARC Rev. Page 44 of 52 --.-~----- c - -. ~, w e-- 1 ~-,w~e,-., - ~ ,~,wa,w--. -s u v -,--w ,.w, . - ~ - - - - ~. -. - - ~
ATTACIIMENT 3 EXPLANATIONS / BASES FOR EALS (Page 21 of 28) '~ EXPLANATIONS / BASES CilART - SAFETY SYSTEM FAILURE OR MALFUNCTION SSFM S. - MODES: 5 & 6 This EAL addresses complete loss of functions required fcr core cooling during refueling and cold shutdown modes. Escalation to Site Area Emergency or General Emergency would be via Abnormal Rad LevelvRadiological Effluent or Duty Emergency Directer / Duty Emergency Manager Judgment, or Fission Product Banier Degradation ICs. This IC and its associated EAL are based on concems raised by Generic Letter 88-17 " Loss of Decay Heat Removal." A mimber of phenomena such as pressurization, vortexing. steam generator U-tube draining. RCS level dilTerences when operating at a mid-loop condition decay heat removal system design. and level instrumentation problems can lead to conditions uhere decay heat reraoval is lost and core uncovery can occur. NRC analyses show that sequences that can cause core uncovery in 15 to 20 minutes and severe core damage within an hour aller decay heat removal is lost. Under these conditions. RCS integrity is lost and fuel clad integrity is lost or potentially lost, w hich is consistent with a Site Arca Emergency. " Uncontrolled" means that system temperature increase is not the result of planned actions by the plant staff. The EAL guidance related to uncontrolled temperature rise is necessary to preserve the anticipatory pnitosophy of NUREG-0654 for events starting from temperatures much low cr than the cold shutdown temperature limit. l SSFM 9.- MODES: S&6 Under the conditions specified by this IC severe core damage can occur a~t reactor coolant system pressure boundary integrity may not be assured. This 1C covers sequences such as prolonged boiling following loss of decay hcc - noval. Thus, declaration of a Site Arca Emergency is warranted under the conditions specified by the IC. Escalation to a General Emergency is via radiological effluent IC. EPP 01-2.1 NUMARC Rev. Page 45 of 52 II
ATTACin!ENT 3 EXPLANATIONS / BASES FOR EALS (Page 22 of 28) EXPLANATIONS / BASES CHART-ADMINISTRATIVE ADM 1. - MODES I TFdOII 2 Required reporting for ECCS actuations and conditions that could lead to a degraded level of salcty of the plant. ADM 2. - MGDES: 1 TIIROUGli 4 Limiting Conditions of Operation (LCOs) rmuire the plant to be brought to a required shutdown mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may or may not be an cmergency or precursor to a + more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specifications requires a one hour report under 10 CFR 50.72 (b) Non-cmcrgency events. Ec plant is within its safety envelope when being shut down within the allowable action statement time in the Technical SpeciDcations. An immediate Notitication of an Unusual Event is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of an Unusual Event is based on the time at uhich the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addicssed by other System Malf t nction. Ilazards, or Fission Product Barrier Degradation ICs. l l ADM 3. - MODES: 1 TIIROUGil 4 A Containment Breach, by itself, is classified as a Notification of Unusual Event in accordaace with Reg. Guide 1.101. l l EPP 01-2.1 NUMARC Rev. Page 46 of 52 m-.,. .m.. r w -, s, e m i,
y ATTACHMENT 3 EXPLANATIONS / BASES FOR EALS (Page 23 of 28) EXPLANATIONS / BASES CHART - LOSS OF PLANT CONTROL / SECURITY COMPROMISE LPOSC L - 310 DES: ALL This EAL is based on the WCGS Site Security Plan. Security events which do not represent at least a potential degradation in the level of safety of the plant. are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. The plant Protected Arca Boundary is typically that part within the sccurity isolation zonc and is denned in the WCGS security plan. Bomb devices discoverca within the plant Vital Area would result in EAL escalation. LPC/SC 2. - 310 DES: ALL This class of security events represents an escalated threat to plant safety above that contained in the Unusual Event. For the purposes of this IC, a civil distmbance which penetrates the protected area boundary can be considered a hostile force. Intrusion into a tital area by a hostile force will escalate this event to a Site Arca Emergency. LPOSC 3. - 310 DES: ALL This class of security events represents an escalated threat to plant safety above that contained in the Alert IC in that a hostile force has pmgressed from the Protected Area to the Vital Area. LPC/SC 4. - 310 DES: ALL This IC encompasses conditions under which a hostile force has taken physical control of vital area required to reach and maintain safe shutdown. LPOSC 5. - 310 DES: ALL With the Control Room cvacuated, additional support, monitoring and direction tiuuugh the Technical Support Center and/or the Emergency Operations Facility is necessary. Inability to establish plant control from outside the Control Room will escalate this event to a Site Arca Emergency. LPOSC 6. - MODES: ALL Expeditious transfer of safety sy stems has not occurred but fission product barrier damage may not yet be indicated. WCGS time for transfer based on analysis or assessments as to how quickly control must be reestablished without core uncovering and/or core damage. This time should not exceed 15 minutes. In cold shutdown and refueling modes, operator concern is directed toward maintaining core cooling such as is discussed in Generic Letter 88-17, " Loss of Decay Heat Removal." In power operation. hot standby, and hot shutdown modes, operator concern is primarily directed toward maintaining critical safety functions and thereby assuring lission product barrier integrity. Escalation of this event, if appropriate, would be by Fission Product Barrier Degradation, Abnormal Rad Levcis/ Radiological Efnuent, or Duty Emergency Director / Duty Emergency Manager Judgment ICs. I i EPP 01-2.1 NUMARC Rev. Page 47 of 52 s.-.,,.-- r-eeg+-e-w -+5-=-tv<.v-r-t r m r* ~ e- -'e---- er se-- -r a'*= ~*-a =
,e m---o
-e' 2---
ATTACHMENT 3 EXPLANATIONS / BASES FOR EALS (Page 24 of 28) EXPLANATIONS / BASES CHART - FIRE FR 1. - M ODES: ALL The purpose of this IC is to address the magnitude and extent of fires that inay be potentially significant precursors to damage to safety systems. His excludes such items as Dres within administration buildings, waste-basket fires, and other small fires of no safety consequmcc. His IC applies to buildings and areas contiguous to plant vital areas or other significant buildings or areas. The intent of this IC is not to include buildings (i.e., warehouses) or areas that are not contiguous or immediately adjacent to plant vital areas. Verification on..e alarm in this context means those actions taken in the control room to determine that the control room alarm is not spurious. FR 2. - MODES: A
- L Re specified areas contain functions and systems required for the safe shutdown of t'ic plant. This list was obtained from WCGS USAR Table 3.3-1. This will make it easier to dete mine if the fire or explosion is potential affecting one or more redundant trains of safety systems.
Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, Fission Product Banier Degradation, Abnormal Rad Levels / Radiological EfDuent, or Duty Emergency Director / Duty Er.crgency Manager Judgment ICs. EPP 01-2.1 NUMARC Rev. Page 48 of 52 c.
ATTACHhlENT 3 EXPLANATIONS / BASES FOR EALS (Page 25 of 28) EXPLANATIONS / BASES CHART - NATURAL PHENOMENA NP L - MODES: ALL NP 1 was developed on WCGS basis. Damage may be caused to some portions of the site. but should not affect ability of snfety functions to operate. Method of detection can be based on instmmentation, validated by a reliable source, or operator assessment. As defined in the EPRI-sponsored " Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989 a " felt carthquake" is: An carthquake of sufricient intensity such that: (a) the vibratory ground motion is felt at the nuclear plant site and recogr' zed as an carthquake based on a consensus of Control Room operators on duty at the time. and (b) for plants with operable scismic instrumentation, the scismic switches of the plant are activated. At Wolf Creck these scismic switches are set at an acceleration of 0.Olg. J l NP 2. - MODES: ALL NP 2 based on WCGS USAA design basis. Scismic cycnts of this magnitude can cause damage to safety functions. l NP 3. - MODES: ALL An carthquake greater than SSE could place the safety systems in a severcly degraded condition. Equipment can be expected to be exposed to forces greater than design limits. 4 NP 4. - MODES: ALL If visual inspection of plant safety related equipment and strudures indicate a loss of a function needed to reach cold shutdown. then emergency escalation is warranted. NP 5. - MODES: ALL NP 5 is based on WCGS USAR Section 3.3.1.1. Wind loads of this magnitud: can cause damage to safety functions. l NP 6. - MODES: ALL NP 6 is based on the assumption that a tornado striking (touching down) within the protected boundary may have potentially damaged plant stmetures containing functions or systems required for safe shutdown of the plant. If such damage is conErmed visually or by o'her in-plant indications. the event may be escalated to Alert. NP 7. - MODES: Atl, This EAL specifics stmeture containing systems and functions required for a safe shutdown of the plant Thi; tist was obtained from WCGS USAR Table 3.3-1. 4 EPP 01-2.1 NUhfARC Rev. Page 49 of 52 -.-m .u. . m.. m. ... m. m 2 ..-- i. w
ATTACIB1ENT 3 EXPL ANATIONS/ BASES FOR EALS - (Page 26 of 28) EXPLANATIONS / BASES CHART - OTIIER HAZARDS i I Oli 1. - MODES: ALL This IC is based on releases in concentrations within the site boundary that will affect the health of plant personnel or affecting the safe operation of the plant with the plant being within the evacuation area of an offsite event (i c., tanker tmck accident releasing toxic gases, etc.) OII 2. - MODES: ALL This IC is based on gases that have entered a plant structure affecting the safe operation of the plant. This IC applies to buildings i and areas contiguous to plant Vital Arcas or other significant buildings or arcas (i.e., Service Water Pump house). The intent of this IC is not to include buildings (i.e., warehouses) or other areas that are not contiguous or immediately adjacent to plant Vital Arcas. It is appropriate that increased monitoring bc I donc to ascertain w hether consequential damage has occurred. Escalation to a higher emergency class, if appropriate, will be based on System Malfunction. Fission Product Barrier Degradation. Abnormal Rad Levels / Radioactive Ellluent. or Duty Emergency Director / Duty Emergency Manager Judgment ICs. Oil 3. - NODES: ALL This EAL is intended to address such items as plane or helicopter crash, or on some sites, train crash, or barge crash that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant. If the crash is confirmed to affect a plant vital area. the event may be escalated to Alert. Oli 4. - MODES: ALL Dis EAL specifics structurcs containing systems and functions required for safe shutdown of the plant This list was obtained from WCGS USAR Table 3.3-1, Oli 5. - MODES: ALL For this EAL only those explosions of sufficient force to damage permanent structures or equipment within the protected area should be considered. As used here, an explosion is a rapid, violent, unconfined combustion. or a catastmphic failure or pressurized equipment, that potentially imparts significant energy to near-by structures and materials. No attempt is made in this EAL to assess the actual magnitude of the damage. The occurrence of the explosion with reports of evidence of damage (e.g., deformation scorching) is sufficient for declaration. The Duty Emergency Director / Duty Emergency Manager also needs to consider any security aspects of the explosion if applicable. l t EPP 01-2.1 NUMARC Rev. Page 50 of 52
ATTACllMENT 3 EXPLANATIONS / BASES FOR EALS (Page 27 of 28) EXPLANATIONS / BASES CHART - OTilER HAZARDS OII 6. - MODES: ALL T - EAL is intended to address main turbine rotating component failures of sufficient magnitud,: to cause observable damage to the turbine casing or to the seals of the turbine generator. Of major concern is the potential for leakage of combustible 11uids (lubricating oils) and gases (hydrogen cooling) to the plant environs. Actual fires and flammable gas build up are appropriately classified under Oli l or the Fire IC for Emergency Clr sification. This EAL is consistent with the definition of an Unusual Event u hile maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment. Escalation of the emergency classilication is based on potential damage done by missiles generated by the failure, or in conjunction w ith a steam generator tube mpture. These latter events would be classified by the radiological ICs or Fission Product Barrier ICs. l OII 7. - MODES: ALL Train derailment that could involve the shipment of radioactive material or cuipment and possible new or spent fuel. l I l Oli H. - MODES: ALL Required notification of transport of a contaminated injured individual to an ofTsite hospital. l 0119. - MODES: ALL This EAL is intended to address unanticipated conditions not addressed explicitly elscu here but that warrant declaration of an cmergency because conditions exist which are believed by the Duty Emergency Director / Duty Emergency Manager to fall under the Unusual Event emergency class. From the broad perspective, one area that may warrant Duty Emergency Director / Duty Emergency Managerjudgment is related to likely or actual breakdown
- of site specific mitigating actions. Examples to consider include inadequate cmergency response procedures, transient response either tmexpected er =
understood, failure or unavailability of emergency systems during an accident in excess of that assumed in accident analysis, or insufficient avatlability of equipment and/or support personnel. Specific example of actual events that may require Duty Emergency Director / Duty Emergency Managerjudgment for Unusual Etent declaration are listed here for consideration. However, this list is by no means all inclusive and is not ictended to limit the discretion of the site to provide fmther examples. . Aircraft crash on-site. . Train derailment on-site. .Near-site explosion which may adversely alTect normal site activities. .Near-site release of toxic or flammable gas which may adversely affect normal site activitics. ~ .Uncontrulled RCS Cooldown due to Secondary Depressurization. It is also intended that the D~uty Emcrgency Directors / Duty Emergency Managersjudgment not be limited by any list of events as defined here or as augmented by the site. Tids list is provided solely as examples for consideration and it is recognized that actual events may not always follow a pre-conceived description. EPP 01-2.1 NUMARC Rev. Page 51 of 52 c
ATTACHMENT 3 EXPLANATIONS / BASES FOR EALS (Page 28 of 28) EXPLANATIONS / BASES CHART - OTHER HAZARDS Oli 10. - MODES: ALL This EAL is intended to address unanticipated conditions not addressed caplicitly elsenhere but that warrant declaration of an emergency because conditions exist which are believed by the Duty Emergency Director / Duty Emergency Manager to fall under the Alert emergency class. OII I1. - MODES: ALL This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist uhich are believed by the Duty Emergency Director / Duty Emergency Manager to fall under the General Emergency class. Oll 12. - MODES: ALL This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are belicted by the Duty Emergency Director / Duty Emergency Manager to fall under the Site Area Emergency class. i EPP 01-2.1 NUMARC Rev. Pagc 52 of 52 . ~ - -. - - - - -.}}