Text

Revised Emergency Plan Implementing Procedures for Catawba, Units 1 & 2
	ML020770296
Person / Time
Site:	Catawba
Issue date:	03/04/2002
From:	Gordon Peterson; Duke Power Co
To:	; Document Control Desk, Office of Nuclear Security and Incident Response
References
	Download: ML020770296 (53)
	v • d • e

P Duke Duke Power OrPower 4800 Concord Rd.

A Duke Energ Cy an rk, SC 29710 (803) 831-4251 OFFICE (803) 831-3221 FAx Gary R. Peterson grpeters@duke-energ. corn Vice President Catawba Nuclear Station March 4, 2002 U.S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001

Subject:

Duke Energy Corporation Catawba Nuclear Station Units 1 and 2 Docket Nos.

50-413 and 50-414 Emergency Plan Implementing Procedures Please find enclosed for NRC Staff use and review the following Emergency Plan Implementing Procedures:

RP/0/A/5000/001, Classification of Emergency (Rev.

015)

RP/0/A/5000/015, Core Damage Assessment (Rev.

005)

These revisions are being submitted in accordance with 10CFR 50.54(q) and do not decrease the effectiveness of the Emergency Plan Implementing Procedures or the Emergency Plan.

By copy of this letter, two copies of the above documents are being provided to the NRC, Region II.

If there are any questions, please call Tom Beadle at 803-831 4027.

Very rly yo s,

Gary R. Peterson Attachments

U.S. Nuclear Regulatory Commission March 4, 2002 Page 2 xc (w/attachments):

L.

A. Reyes U.S. Nuclear Regulatory Commission Regional Administrator, Region II Atlanta Federal Center 61 Forsyth St.,

SW, Suite 23T85

Atlanta, GA 30303 (w/o attachments):

C.

P. Patel NRC Senior Project Manager (CNS)

U.S. Nuclear Regulatory Commission Mail Stop 0-8 H12 Washington, DC 20555-0001 D.

J.

Roberts Senior Resident Inspector (CNS)

U.S. Nuclear Regulatory Commission Catawba Nuclear Site

PROCEDURE RP/0/A/5000/001 RP/0/A/5000/002 RP/O/AI5000/003 RP/O/A/5000/004 RP/O/A15000/005 RP/0/A/5000/06 RP/0/A/5000/006 A RP/O/A/5000/006 B RP/0/A/5000/006 C RP/0/A/5000/007 RP/O/A/5000/08 RP/O/B/5000/008 RP/O/A/5000/009 RP/O/A/5000/0 10 RP/O/A/5000/l 1 RP/O/B/5000/ 12 RP/OIB/5000/0 13 RP/O/B/5000/14 RP/0/A15000/015 RP/O/B/5000/0 16 RP/O/B/5000/17 Page 1 of 4 DUKE POWER COMPANY CATAWBA NUCLEAR STATION EMERGENCY PLAN IMPLEMENTING PROCEDURES INDEX VOLUME I TITLE Classification of Emergency (Rev. 015)

Notification of Unusual Event (Rev. 035)

Alert (Rev. 037)

Site Area Emergency (Rev. 039)

General Emergency (Rev. 039)

Deleted Notifications to States and Counties from the Control Room (Rev. 014)

Notifications to States and Counties from the Technical Support Center (Rev. 014)

Deleted Natural Disaster and Earthquake (Rev. 021)

Deleted Spill Response (Rev. 020)

Collision/Explosion (Rev. 006)

Conducting A Site Assembly or Preparing the Site for an Evacuation (Rev. 014)

Deleted Deleted NRC Notification Requirements (Rev. 029)

Deleted Core Damage Assessment (Rev. 005)

Deleted Deleted February 28, 2002

Page 2 of 4 PROCEDURE RP/O/A/5000/0 18 RP/0/B/5000/019 RP/O/A/5000/020 RP/O/A/5000/021 RP/0/B/5000/022 RP/0/B/5000/023 RP/0/A/5000/024 RP/0/B/5000/025 RP/O/B/5000/026 "RP/0/B/5000/028 DUKE POWER COMPANY CATAWBA NUCLEAR STATION EMERGENCY PLAN IMPLEMENTING PROCEDURES INDEX VOLUME I TITLE Emergency Worker Dose Extension (1/15/96)

Deleted Technical Support Center (TSC) Activation Procedure (Rev. 015)

Deleted Evacuation Coordinator Procedure (Rev. 004)

Deleted OSC Activation Procedure (Rev. 009)

Recovery and Reentry Procedure (Rev. 003)

Site Response to Security Events (Rev. 003)

Communications and Community Relations EnergyQuest Emergency Response Plan (Rev. 001)

February 28, 2002

Page 3 of 4 PROCEDURE HP/0/B/1000/006 HP/0/B/1009/001 HP/0/B/1009/003 HP/0/B/1009/004 HP/0/B/1009/005 HP/0/B/1009/006 HP/0/B/1009/007 HP/0/B/1009/008 HP/0/B/1009/009 HP/0/B/1009/014 HP/0/B/1009/016 HP/0/B/1009/017 HP/1/B/1009/017 HP/2/B/1009/017 HP/0/B/1009/018 HP/0/B/1009/019 HP/O/B/1009/024 DUKE POWER COMPANY CATAWBA NUCLEAR STATION EMERGENCY PLAN IMPLEMENTING PROCEDURES INDEX VOLUME II TITLE Emergency Equipment Functional Check and Inventory (Rev. 053)

Radiation Protection Recovery Plan (Rev. 008)

Radiation Protection Response Following a Primary to Secondary Leak (Rev. 008)

Environmental Monitoring for Emergency Conditions Within the Ten-Mile Radius of CNS (Rev. 028)

Personnel/Vehicle Monitoring for Emergency Conditions (Rev. 016)

Alternative Method for Determining Dose Rate Within the Reactor Building (Rev. 008)

In-Plant Particulate and Iodine Monitoring Under Accident Conditions (Rev. 019)

Contamination Control of Injured Individuals (Rev. 015)

Guidelines for Accident and Emergency Response (Rev. 039)

Radiation Protection Actions Following an Uncontrolled Release of Radioactive Material (Rev. 008)

Distribution of Potassium Iodide Tablets in the Event of a Radioiodine Release (Rev. 011)

Deleted Deleted Deleted Deleted Emergency Radio System Operation, Maintenance and Communication (Rev. 010)

Implementing Procedure for Estimating Food Chain Doses Under Post Accident Conditions (Rev. 002)

February 28, 2002

Page 4 of 4 PROCEDURE HP/O/B/1009/025 HPlO/B!1009/026 SH/0/B/2005/001 SH/O/B/2005/002 OP/O/A/6200/021 SR/0/B/2000/001 SR/OfB/2000/002 SR/0/B/2000/003 SR/O/B/2000/004 DUKE POWER COMPANY CATAWBA NUCLEAR STATION EMERGENCY PLAN IMPLEMENTING PROCEDURES INDEX VOLUME II TITLE Deleted On-Shift Offsite Dose Projections (Rev. 003)

Emergency Response Offsite Dose Projections (Rev. 001)

Protocol for the Field Monitoring Coordinator During Emergency Conditions (Rev. 002)

Post Accident Liquid Sampling System II+ (Rev. 034)

Standard Procedure for Public Affairs Response to the Emergency Operations Facility (Rev. 003)

Standard Procedure for EOF Services (Rev. 002)

Activation of the Emergency Operations Facility (Rev. 008)

Notification to States and Counties from the Emergency Operations Facility (Rev. 004)

February 28, 2002

(l1O4-M1 Duke Power Company PROCEDURE PROCESS RECORD (1)ID No. RP/O/N5o0o/015 Revision No.

5 3AD AT'rllT 2

Station CATAWBA NUCLEAR STATION (3)

Procedure Title (4) Prepared By Core Damaae Assessment Date

-214 z (5) Requires NSD 228 Applicability Determination?

El Yes (New procedure or revision with major changes)

ONo (Revision with minor changes)

U No (To incorporate previously approved changes)

(6)

Reviewed By --

/I/*.1-

_za4 Cross-Disciplinary Review By Reactivity Mgmt. Review By Mgmt. Involvement Review By (7)

Additional Reviews Reviewed By 64-1-4 L-(QR)

Date Z

--z i -

6 (QR) NA _4 Date.

Lz (OR) NA __Date__

(Ops. Supt.) NA_!§L__Date. 2-/2 1/k2,,

(P Date Z-

-Z*O

"-ef Date Reviewed By (r

Temporary Approval (if necessary)

"By IOSM/QR) Date By (9) Approved By QR) Date (V-a j Date -- 2_'- o2z.

PERFORMANCE (Compare with control copy every 14 calendar days while work is being performed.)

(10) Compared with Control Copy Compared with Control Copy Compared with Control Copy Date (11) Date(s) Performed Work Order Number (WO#)

COMPLETION (12) Procedure Completion Verification

[] Yes 0 N/A Check lists and/or blanks properly initialed, signed, dated, or filled in N/A, as appropriate?

O Yes 0l N/A Required enclosures attached?

O Yes El N/A Data sheets attached, completed, dated, and signed?

0I Yes 0l N/A Charts, graphs, etc. attached, dated, identified, and marked?

El Yes 0l N/A Procedure requirements met?

Verified By fl tc

(.

-rocedure Completion Approved Date (14) Remarks (attach additional pages, if necessary)

Date n

+

Verified By Daten L.

'JJA" Bv t'l I

Duke Power Company Catawba Nuclear Station Core Damage Assessment

--E.P.I.P-Continuous Use Procedure No.

RP/O/A/5000/015 Revision No.

005 Electronic Reference No.

CN005GNY J ______________________________________________________________

I

RP/O/A/5000/015 Page 2 of 12 Core Damage Assessment

1. Symptoms NOTE:

This procedure will normally be performed by Nuclear Engineers while in the Technical Support Center (TSC) to provide a means of determining the status of the core based on various parameters.

1.1 1(2) EME 53, "Containment Radiation Monitor" in alarm.

1.2 High Core Exit Thermocouple (CET) readings.

1.3 Low Reactor Vessel Level Indication System (RVLIS) levels.

1.4 High containment hydrogen concentrations.

1.5 Any condition in which failed fuel is suspected.

2. Immediate Actions None

3. Subsequent Actions 3.1 Identify Current Plant Status 3.1.1 Complete the following table based on current plant data.

Time of data, (mm/dd/yy hh:mm)

Time of Reactor Shutdown, To Hours Core Exit Thermocouples (CET)

(GD ERORXG, P0828)

Deg F RVLIS (GD ERORXG, P0180 or P0181)

Containment Radiation Monitors 1(2)EMF 53A or B R/hr (GD ERORXG, A1308 or A1314)

RP/0/A/5000/015 Page 3 of 12 Determine possible status of reactor core using table below:

3.1.2

RP/O/A/5000/015 Page 4 of 12 Figure 1 Containment Radiation Level vs. Time for RCS Release 100 200 300 400 500 600 700 Time After Shutdown, hours Time After Containment Dose Rate Shutdown (hrs)

(Rad/hr) 0.5 9.1808 1

8.8621 2

8.3792 8

7.0574 16 6.2611 24 5.7672 100 3.8545 240 2.3002 720 0.41169 10 cc 0

1 E

0

0.

0.1 0

800 Containment Radiation Level vs.

Ig Time for RCS Release

RPIOtAI50001015 Page 5 of 12 Figure 2 Containment Radiation Level vs. Time for 1% Fuel Overtemperature Release

- -

i -

-I -I-----f- -I -

4 -= -,

i -

4 -

4- -I -

4 -

.1 -

I -

4 -

I jj71

Ii

- I -

600 r

4.!

L4 4

1 141~

l f

1.OOE+04

,. 1.00E+03 "OA o 1.00E+02 E

0 0

1.00E+01 200 300 4-4-.

400 Ji.

500 700 Time After Shutdown, hours RCS Press < 1600 psig, no spray 4-RCS Press < 1600 psig, withpay Pe >1r-RSPress>

1600 pg spray

ý:RCS Press > 1600 psig, no spray -"-*-RCS Press > 1600 psig, with sprayl Fure 2

Containment Radiation Level vs. Time for 1 % Fuel Overtemperature Release Time After RCS Pressure RCS Pressure RCS Pressure RCS Pressure Shutdown

< 1600 psig,

> 1600 psig,

> 1600 psig, (hrs) no Spray with Spray no Spray with Spray (Rad/Hr)

(Rad/Hr)

-(Rad/Hr (Rad/Hr) 0.5 5.30E+03 5.07E+03 2.23E+03 2.21 E+03 1

4.39E+03 4.20E+03 1.85E+03 1.84E+03 2

3.16E+03 3.02E+03 1.32E+03 1.31 E+03 8

1.13E+03 1.06E+03 4.32E+02 4.29E+02 16 6.45E+02 5.97E+02 2.30E+02 2.27E+02 24 5.15E+02 4.79E+02 1.84E+02 1.82E+02 100 2.90E+02 2.76E+02 1.08E+02 1.07E+02 240 1.54E+02 1.45E+02 5.11E+01 5.06E+01 720 3.49E+01 3.13E+01 5.44E+00 5.26E+00 I

U' ThTh

I

'# L LII:>

1.OOE+00 ifi

44

i-I L4

0 100 800 I

I IIIIIIIIIIIIIIIIIlIII 7*

III1*111 i

I S..

I I

T I

I I

]

I

-1ý 1 M'rft-'ftM6..

m C

RP/O/A/5000/0 15 Page 6 of 12 3.1.3 IF status of core is "no core damage", exit this procedure and continue to monitor plant conditions. If conditions warrant, re-run this procedure.

IF status of core is "possible fuel rod cladding damage", proceed to Step 3.2.

IF status of core is "possible fuel overtemperature damage", proceed to Step 3.3.

3.2 Clad Damage Assessment NOTE:

EMF 53 may not be useful to assess core damage for containment bypass sequences (e.g.,

S/G tube ruptures).

3.2.1 Record 1(2)EMF 53 reading and other data:

R/hr at hours after shutdown.

Containment Spray:

LI ON EL OFF RCS Pressure:

psig 3.2.2 Determine "Predicted Containment Radiation Level at 100% Clad Damage" using Figure 3.

Predicted Containment Radiation Level at 100% Clad Damage:

R/Hr 3.2.3.

Estimate clad damage:

% Clad DamagecRM = (EMF 53 + Predicted Cont. Rad Level at 100%)

100

% Clad DamagecRM = (

+

)*100=

RP/O/A/5000/015 Page 7 of 12 Figure 3 Containment Radiation Level vs. Time for 100% Clad Damage Release ft IA I f

-I 1.OOE+05 1.OOE+04 6 1.OOE+03 M

0 E 1.00E+02 0

0 100

.4 I-

1-f<, t -

+ 4 4 = 4 44Zz44 200

=

1

-4 300 400 500 600 Time After Shutdown, hours RCS Press < 1600 psig, no spray ----

RCS Press < 1600psig. with spray RCS Press> 1600 psig, no spray -RCS Press> 1600 psig, with spray 0

F 4-4 i

1.OOE+01 1.OOE+00

-MMM 700 800 Figure 3 Containment Radiation Level vs. Time for 100% Clad Damage Release Time After RCS Pressure RCS Pressure RCS Pressure RCS Pressure Shutdown

< 1600 psig,

> 1600 psig,

> 1600 psig, (hrs) no Spray with Spray no Spray with Spray (Rad/Hr)

(Rad/Hr)

(Rad/Hr) 0.5 3.37E+04 2.91E+04 1.11 E+04 1.1OE+04 1

2.78E+04 2.42E+04 9.24E+03 9.17E+03 2

2.03E+04 1.75E+04 6.58E+03 6.52E+03 8

8.09E+03 6.25E+03 2.17E+03 2.13E+03 16 4.96E+03 3.48E+03 1.15E+03 1.12E+03 24 3.98E+03 2.68E+03 9.24E+02 8.98E+02 100 2.19E+03 1.30E+03 5.40E+02 5.22E+02 240 1.29E+03 6.23E+02 2.57E+02 2.44E+02 720 4.22 E+02 6.54E+01 2.81E+01 2.10E+01 I--

. =:I I

I I

I I--

I I

I -

I 7ý7 ý111 ý!I I

RP/0/A/5000/015 Page 8 of 12

-3.2.4 Record number of available core exit thermocouples(CETs)(GD EROCOREI(2,3)):

3.2.5 IF RCS pressure is greater than 1600 psig, record number of CETs greater than 1400 TF:

3.2.6 IF RCS pressure is less than 1600 psig, record number of CETs greater than 1200 TF:

-3.2.7 Estimate clad damage:

% Clad DamagecET = (3.2.5 OR 3.2.6 ÷ 3.2.4)

100

% Clad DamagecET = (

+ _)*

100=

.3.2.8 Confirm reasonableness of clad damage estimates using expected responses below:

"* RVLIS less than 54% AND greater than 39%

"* Hot Leg RTD greater than Tsar AND less than 650TF

"* Source Ranges greater than 104 cps

"* Difference in clad damage estimates from Containment Radiation Monitor (EMIF 53) and CETs less than 50%, using:

ABSF %CladDamagecRM - %CladDamagec1 L

%Cl

%Cladna amagecJ IL

%CladDamagecRM 3.2.9 IF the expected response is not obtained, determine if the deviation can be explained from the accident progression

"* Injection of water to the RCS

"* Bleed Paths from the RCS

"* Direct radiation to the containment radiation monitors OR from conservatism in the predictive model fuel burnup fission product retention in the RCS fission product removal from containment

RP/O/A/5000/015 Page 9 of 12 3.2.10 Report estimate of clad damage to TSC Engineering Manager and to EOF Reactor Physics.

3.3 Fuel Overtemperature Damage Assessment 3.3.1 Record 1(2)EMF 53 reading and other data:

R/hr at hours after shutdown.

Containment Spray:

11 ON 11 OFF RCS Pressure:

psig 3.3.2 Determine "Predicted Containment Radiation Level at 100% Overtemp Damage" using Figure 4.

Predicted Containment Radiation Level at 100% Overtemp Clad Damage:

R/Hr

_ 3.3.3 Estimate core damage:

% Core DamagecRM = (EMF 53 + Predicted Cont. Rad Level at 100%)

100

% Core DamageCRM=(

=

)* 100 =

RP/O/A/5000/015 Page 10 of 12 Figure 4 Containment Radiation Level vs. Time for 100% Fuel Overtemperature Release 1.OOE+06 o 1.OOE+05 "U

{0 o 1.00E+04 a

C E

c 0

T Th

'Th I

I

>4

$-lA-I

Li4

-4 j

4

'I*

-I 100 200 300 U

-4 v 400

1 500 600 I-700 Time After Shutdown, hours P--0RCS Press < 1600 psig, no spray --

RCS Press < 1600 psig, with sprayl

--*f-RCS Press> 1600 psig, no spray -- '-RCS Press> 1600 psig, with sprayj k-I.UU00-HU 1.OOE+02 0

d I 800 Containment Radiation Level vs. Time for 100% Fuel Overtemperature Release Time RCS Pressure RCS Pressure RCS Pressure RCS Pressure After

< 1600 psig,

> 1600 psig,

> 1600 psig, Shutdown no Spray with Spray no Spray with Spray (hrs)

(Rad/Hr)

(Rad/Hr) 0.5 5.30E+05 5.07E+05 2.23E+05 2.21 E+05 1

4.39 E+05 4.20E+05 1.85E+05 1.84E+05 2

3.16E+05 3.02E+05 1.32E+05 1.31 E+05 8

1.13E+05 1.06E+05 4.32E+04 4.29E+04 16 6.45E+04 5.97E+04 2.30E+04 2.27E+04 24 5.15E+04 4.79E+04 1.84E+04 1.82E+04 100 2.90E+04 2.76E+04 1.08E+04 1.07E+04 240 1.54E+04 1.45E+04 5.11E+03 5.06E+03 720 3.49E+03 3.13E+03 5.44E+02 5.26E+02 11111 i

I 4 -

-4.

I

- Ft

?

RP/0/A/5000/015 Page 11 of 12 3.3.9 from conservatism in the predictive model

"* fuel burnup

"* fission product retention in the RCS fission product removal from containment Report estimate of core damage to TSC Engineering Manager and to EOF Reactor Physics.

3.3.4 Record number of available core exit thermocouples(CETs)(GD EROGROUP 1(2,3)): _

3.3.5 Record number of CETs greater than 2000 'F:

3.3.6 Estimate core damage:

% Core DamagecEr = (3.3.5 - 3.3.4)

100

% Core DamagecEr = (

-)*

100 =

3.3.7 Confirm reasonableness of core damage estimates using expected responses below.

RVLIS less than 39%

Hot Leg RTD greater than 650TF Source Ranges greater than 104 cps Difference in core damage estimates from Containment Radiation Monitors (CRM) and CET's less than 50%, using:

ABS[ % Core DamagecRI - % Core Damagecz

% Core DamagecRM j

3.3.8 IF the expected response is not obtained, determine if the deviation can be explained from the accident progression

Injection of water to the RCS Bleed Paths from the RCS Direct radiation to the containment radiation monitors OR

RP/O/A/5000/015 Page 12 of 12

4. Enclosures None

5. References 5.1 Duke Power Calculation DPC-1229.00-00-0006, "Estimated Sample Dose and CDAG Setpoint Calculations in Support of PASS Removal", Rev 0, 6/01/01.

5.2 WCAP-14696-A, " Westinghouse Owners Group Core Damage Assessment Guidance",

Revision 1, November, 1999.

5.3 Unit Data Book.

5.4 MNS procedure RP/O/A/5700/019, Core Damage Assessment.

(R04-OI)

Duke Power Company PROCEDURE PROCESS RECORD (1) ID No.

RP/OA/5000/001 Revision No.

015 P'

"'ARATION (2)-Station Catawba (3)

Procedure Title CIlassification of Emergencv (4)

Prepared By ea-ff-(5)

Requires NSD 228 Applicability Determination?

[]

Yes (New procedure or revision with major changes)

El No (Revision with minor changes) li No (To incorporate previously approved changes (6)

Reviewed By (QR)

Cross-Disciplinary Review By (QR)

NA Reactivity Mgmt. Review By __(QR)

NA__

Mgmt. Involvement Review By (Ons. Sunt.

NA&""

(7)

Additional Reviews Reviewed By Reviewed By Date Z'z c*2-Date Date Date Date Z- -11 Date Date (8)

Temporary Approval (if necessary)

By By OSM/QR)

(i Approved By Date PERFORMANCE (Compare with control copy every 14 calendar days while work is being performed.)

(10)

Compared with Control Copy Date Compared with Control Copy Date Compared with Control Copy Date Date

-Z) 2T /0 (I1)

Date(s) Performed Work Order Number (WO#)

COMPLETION (12)

Procedure Completion Verification:

El Yes El NA Check lists and/or blanks initialed, signed, dated, or filled in NA. as appropriate?

El Yes El NA Required enclosures attached?

El Yes El NA Data sheets attached, completed, dated, and signed" El Yes El NA Charts, graphs. etc. attached, dated, identified, and marked?

El Yes El NA Procedure requirements met?

Verified By (13)

Procedure Completion Approved (14)

Remarks (Attach additional pages, if necessary)

Date Date

Duke Power Company Catawba Nuclear Station Classification of Emergency Multiple Use L

Procedure No.

RPIO/A/5000/001 Revision No.

015 Electronic Reference No.

CN005GNK CNOO5GNK

RP/O/A/5000/001 Page 2 of 3 Classification of Emergency

1. Symptoms 1.1 Notification of Unusual Event 1.1.1 Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant.

1.1.2 No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety occurs.

1.2 Alert 1.2.1 Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant.

1.2.2 Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

1.3 Site Area Emergency 1.3.1 Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public.

1.3.2 Any releases are not expected to exceed EPA Protective Action Guideline exposure levels except near the site boundary.

1.4 General Emergency 1.4.1 Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity.

1.4.2 Releases can be reasonably expected to exceed EPA Protective Action Guidelines exposure levels offsite for more than the immediate site area.

2. Immediate Actions 2.1 Determine operating mode that existed at the time the event occurred prior to any protection system or operator action initiated in response of the event.

2.2 IF the plant was in Mode 1-4 and a valid condition affects fission plrodLuct barriers, proceed to Enclosure 4. 1.

RP/O/A/5000/001 Page 3 of 3 2.3 IF a General Emergency is NOT declared in Step 2.2 OR the condition does not affect fission product barriers, review the listing of enclosures to determine if the event is applicable to one the categories shown.

2.4 Compare actual plant conditions to the Emergency Action Levels listed, then declare the appropriate Emergency Class as indicated.

2.5 Implement the applicable Emergency Response Procedure (RP) for that classification and continue with subsequent steps of this procedure.

Notification of Unusual Event RP/O/A/5000/002 Alert RP/O/A/5000/003 Site Area Emergency RP/O/A/5000/004 General Emergency RP/O/A/5000/005

3.

Subsequent Actions 3.1 To escalate, de-escalate, or terminate the Emergency, compare plant conditions to the Initiating Conditions of Enclosures 4.1 through 4.7.

3.2 Refer to enclosure 4.9, Emergency Declaration Guidelines, as needed.

4. Enclosures 4.1 Fission Product Barrier Matrix 4.2 System Malfunctions 4.3 Abnormal Rad Levels/Radiological Effluent 4.4 Loss of Shutdown Functions 4.5 Loss of Power 4.6 Fires/Explosions and Security Events 4.7 Natural Disasters, Hazards and Other conditions Affecting Plant Safety 4.8 Definitions/Acronyms 4.9 Emergency Declaration Guidelines 4.10 Radiation Monitor Reading for Enclosure 4.3 EALs

Enclo(

ý 4.1 Fission Barrier Matrix RP/O/A/I

'001 Page 1 of 5 Use EALs to determine Fission Product Barrier status (Intact, Potential Loss, or Loss). Add points for all 3 barriers. Classify according to the table below.

Note 1: This table is only applicable in Modes 1-4.

Note 2: Also, an event (or multiple events) could occur which results in the conclusion that exceeding the Loss or Potential Loss thresholds is IMMINENT (i.e., within 1-3 hours). In this IMMINENT LOSS situation, use judgement and classify as if the thresholds are exceeded.

Note 3: When determining Fission Product Barrier status, the Fuel Clad Barrier should be considered to be lost or potentially lost if the conditions for the Fuel Clad Barrier loss or potential loss EALs were met previously during the event, even if the conditions do not currently exist.

Note 4: Critical Safety Function (CSF) indications are not meant to include transient alarm conditions which may appear during the start-up of engineered safeguards equipment. A CSF condition is satisfied when the alarmed state is valid and sustained. The STA should be consulted to affirm that a CSF has been validated and the appropriate functional restoration procedure has been implemented prior to the CSF being used as a basis to classify an emergency.

EAL #

Unusual Event EAL #

Alert EAL #

Site Area Emergency EAL #

General Emergency

4. I.UI Potential Loss of

4. I.A.I Loss OR Potential Loss 4.1.S.1 Loss OR Potential Loss 4.I.G.1 Loss of All Three Barriers Containment of of Both Nuclear Coolant System Nuclear Coolant System AND Fuel Clad

4. 1.U.2 Loss of Containment

4. I.A.2 Loss OR Potential Loss 4.1.S.2 Loss

4. I.G.2 Loss of Any Two Barriers o f AND AND Fuel Clad Potential Loss Potential Loss of the Third Combinations of Both Nuclear Coolant System AND Fuel Clad

4. I.A.3 Potential Loss of 4.1.S.3 Loss of Containment Containment AND AND Loss OR Potential Loss Loss OR Potential Loss of Any Other Barrier of Any Other Barrier

RP/O/A/5f

'001 Page 2 of 5 Enclon B..

e Ma 4.1 Fission Barrier Matrix NOTE: If a barrier is affected, it has a single point value based on a "potential loss" or a "loss". "Not Applicable" is included in the table as a place holder only, and has no point value assigned.

Barrier Points (1-5)

Potential Loss (X)

Loss (X)

Total Points Classification Containment 1

3 1 - 3 Unusual Event NCS 4

5 4-6 Alert Fuel Clad 4

5 7-10 Site Area Emergency Total Points 11 -13 General Emergency

1. Compare plant conditions against the Fission Banrier Matrix on pages 3 through 6 of 6.

2. Determine the "potential loss" or "loss" status for each barrier (Containment, NCS and Fuel Clad) based on the EAL symptom description.

3. For each barrier, write the highest single point value applicable for the barrier in the "Points" column and mark the appropriate "loss" column.

4. Add the points in the "Points" column and record the sum as "Total Points".

5. Determine the classification level based on the number of "Total Points".

6. In the table on page I of 6, under the "classification" column, select the event number (e.g. 4. I.A. I for Loss of Nuclear Coolant System) that best fits the loss of barrier descriptions.

7. Using the number (e.g. 4.1.A. 1) select the preprinted notification form and complete the required information for Emergency Coordinator approval and transmittal.

(

EncIos"e 4.1 Fission Batk. er Matrix RP/O/A/5(

too I Page 3 of -

4.1.C CONTAINMENT BARRIER POTENTIAL LOSS (I Point)

LOSS (3 Points)

1. Critical Safety Function Status Containment-RED Not applicable Core cooling-RED Path is indicated for >15 minutes

2. Containment Conditions Containment Pressure > 15 PSIG H2 concentration >

9%

Containment pressure greater than 3 psig with less than one full train of NS and a VX-CARF operating.

Rapid unexplained decrease in containment pressure following initial increase Containment pressure or suMp level response not consistent with LOCA conditions.

Ip 4.1.N NCS BARRIER POTENTIAL LOSS (4 Points)

LOSS (5 Points)

1. Critical Safety Function Status NCS Integrity-Red Not applicable Heat Sink-Red

2. NCS Leak Rate Unisolable leak exceeding the capacity of one charging pump in the normal charging mode with letdown isolated.

GREATER THAN available makeup capacity as indicated by a loss of NCS subcooling.

4.1.F FUEL CLAE POTENTIAL LOSS (4 Points)

BARRIER LOSS (5 Points)

1. Critical Safety Function Status Core Cooling Orange Core Cooling-Red 6

Heat Sink-Red

2. Primary Coolant Activity Level 6

Not applicable Coolant Activity GREATER THAN 300 ILCi/cc Dose Equivalent Iodine (DEI) 1-131 CONTINUED C

CONTINUED CONTINUED

Enckr 4.1 RPIO/A/4 Fission Barrier Matrix Page 4 of 5 4.1.C CONTAINM ENT BARRIER POTENTIAL LOSS (I Point)

LOSS (3 Points)

3. Containment Isolation Valves Status After Containment Isolation Actuation Not applicable Containment isolation is incomplete and a release path from containment exists

4. SG Secondary Side Release With Primary-to-Release of secondary side to the environment with primary to secondary leakage GREATER THAN Tech Spec allowable 4.1.N NCS BARRIER POTENTIAL LOSS (4 Points)

LOSS (5 Points)

II im

3. SG Tube Rupture Primary-to Secondary leak rate exceeds the capacity of one chargino pump in the normal charging mode with letdown isolated.

Indication that a SG is Ruptured and has a Non-Isolable secondary line fault Indication that a SG is ruptured and a prolonged release of contaminated secondary coolant is occurring from the affected SG to the environment

4. Containment Radiation Monitoring 0

Not applicable 0

Not applicable CONTINUED 4.1.F POTENTIAL LOSS (4 Points)

LOSS (5 Points)

3. Containment Radiation Monitoring Not applicable Containment radiation monitor 53 A or 53 B reading >117 R/hr

4. Emergencv Coordinator/EOF Director Judgement Any condition, including inability to monitor the barrier, that in the opinion of the Emergency Coordinator/EOF Director indicates LOSS or POTENTIAL LOSS of the fuel clad barrier.

END

(

'001 FUEL CLAD BARRIER Secondary Leakage Not applicable

[ I II CONTINUED

Enclo7,'"e 4.1 Fission Bait.er Matrix RP/O/A/5

%'/O01 Page 5 of 4.1.C CONTAINMENT BARRIER POTENTIAL LOSS (I Point)

LOSS (3 Points)

5. Significant Radioactive Inventory In Containment Containment Rad.

Monitor EMF53A or 53B Reading @ time since shutdown:

>470 R/hr @ 0-0.5 hr

> 170 R/hr @ 0.5-2 hr

> 125 R/hr @ 2-4 hr

> 90 R/hr @ 4 - 8 hr

> 53 R/hr @ > 8 hr 6

Not applicable

6. Emergency Coordinator /EOF Director Judgement Any condition, including inability to monitor the barrier, that in the opinion of the Emergency Coordinator/EOF Director indicates LOSS or POTENTIAL LOSS of the containment barrier.

END 4.1.N NCS BARRIER POTENTIAL LOSS (4 Points)

LOSS (5 Points)

5. Emergency Coordinator/EOF Director Judgement Any condition, including inability to monitor the barrier, that in the opinion of the Emergency Coordinator /EOF Director indicates LOSS or POTENTIAL LOSS of the NCS barrier.

END 4.1.F FUEL CL POTENTIAL LOSS (4 Points)

AD BARRIER LOSS (5 Points)

(

H Imll roll

Encdosur, System Malfunctions RP/O/A/5000/001 Page I of 2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.2.U.1 Inability to Reach Required 4.2.A.1 Unplanned Loss of Most or All Shutdown Within Technical Safety System Annunciation or Specification Limits.

Indication in Control Room With Either (1) a Significant OPERATING MODE:

1, 2, 3, 4 Transient in Progress, or (2)

Compensatory Non-Alarming 4.2.U.1-1 Plant is not brought to required Indicators Unavailable.

operating mode within Technical Specifications LCO Action Statement OPERATING MODE:

1, 2, 3, 4 Time.

4.2.U.2 Unplanned Loss of Most or All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes.

OPERATING MODE:

1,2, 3, 4 4.2.U.2-1 The following conditions exist:

Unplanned loss of most (>50%)

annunciators associated with safety systems for greater than 15 minutes.

AND In the opinion of the Operations Shift Manager/Emergency Coordinator/EOF Director, the loss of the annunciators or indicators requires additional personnel (beyond normal shift compliment) to safely operate the unit.

4.2.A.1-1 The following conditions exist:

Unplanned loss of most (>50%)

annunciators associated with safety systems for greater than 15 minutes.

AND In the opinion of the Operations Shift Manager/Emergency Coordinator/EOF Director, the loss of the annunciators or indicators requires additional personnel (beyond normal shift compliment) to safely operate the unit.

AND EITHER of the following:

A significant plant transient is in progress Loss of the OAC.

4.2.S.1 Inability to Monitor a Significant Transient in Progress.

OPERATING MODE:

1, 2,3,4 4.2.S.1-1 The following conditions exist:

Loss of most (>50%)

Annunciators associated with safety systems.

AND A significant plant transient is in progress.

AND Loss of the OAC.

AND Inability to provide manual monitoring of any of the following Critical Safety Functions:

subcriticality

"* core cooling

"* heat sink

"* containment.

END CONTINUED END C

(

END

(

Enclosur I

RP/O/A/5000/001 System Malfunctions Page 2 of 2 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.2.U.3 Fuel Clad Degradation.

OPERATING MODE:

1, 2, 3:*

4.2.U.3-1 Dose Equivalent 1-131 greater than the Technical Specifications allowable limit. (*Mode 3 with TAV >5000 F) 4.2.U.4 Reactor Coolant System (NCS)

Leakage.

OPERATING MODE:

1, 2,3,4 4.2.U.4-1 Unidentified leakage > 10 gpm.

4.2.U.4-2 Pressure boundary leakage > 10 gpm.

4.2.U.4-3 Identified leakage > 25 gpm 4.2.U.5 Unplanned Loss of All Onsite or Offsite Communications.

OPERATING MODE:

ALL 4.2.U.5-1 Loss of all onsite communications capability (internal phone system, PA system, onsite radio system) affecting the ability to perform routine operations.

4.2.U.5-2 Loss of all offisite communications capability (Selective Signaling, NRC ETS lines, offsite radio system, commercial phone system) affecting the ability to communicate with offsite authorities.

END

Enclosure(

Abnormal Rad Levels/Radiological Effluent RP/O/A/5000K,..

Page 1 of 5 UNUSUAL EVENT 4.3.U.1 Any Unplanned Release of Gaseous 4.3.A.1 or Liquid Radioactivity to the Environment that Exceeds Two Times the SLC Limits for 60 Minutes or Longer.

OPERATING MODE:

ALL 4.3.U.1-1 A valid Trip 2 alarm on radiation OPERAT monitor EMF-49L or EMF-57 for > 60 minutes or will likely continue for > 60 4.3.A.1-1 minutes which indicates that the release may have exceeded the initiating condition and indicates the need to assess the release with procedure HP/O/B/ 1009/014.

4.3.U.1-2 A valid indication on radiation monitor EMF-36L of > 3.OOE+04 cpm for > 60 minutes or will likely continue for Ž- 60 minutes, which indicates that the release may have exceeded the initiating condition and indicates the need to assess the release with procedure SHI/0/B/2005/001.

ALERT Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times the SLC limits for 15 Minutes or Longer.

'ING MODE:

ALL A valid indication on radiation monitor EMF-49L or EMF-57 of > 1.2E+05 cpm for_> 15 minutes or will likely continue for >15 minutes, which indicates that the release may have exceeded the initiating condition and indicates the need to assess the release with procedure HP/O/B/ 1009/014.

(Continued)

SITE AREA EMERGENCY 4.3.S.1 Boundary Dose Resulting from an Actual or Imminent Release of Radioactivity Exceeds 100 mRem TEDE or 500 mRem CDE Adult Thyroid for the Actual or Projected Duration of the Release.

OPERATING MODE:

ALL 4.3.S.1-1 A valid indication on radiation monitor EMF 36L of> 2.7E+06 cpm sustained for_> 15 minutes.

4.3.S.1-2 Dose assessment team calculations indicate dose consequences greater than 100 mRem TEDE or 500 mRem CDE Adult Thyroid at the site boundary.

(Continued)

GENERAL EMERGENCY 4.3.G.1 Boundary Dose Resulting from an Actual or Imminent Release of Radioactivity that Exceeds 1000 mRem TEDE or 5000 mRem CDE Adult Thyroid for the Actual or Projected Duration of the Release.

OPERATING MODE:

ALL 4.3.G.1-1 A valid indication on radiation monitor EMF 36H of> 8.3E+03 cpm sustained for> 15 minutes.

4.3.G.1-2 Dose assessment team calculations indicate dose consequences greater than 1000 mRem TEDE or 5000 mRem CDE Adult Thyroid at the site boundary.

(Continued)

(

Enclosur 4 Abnormal Rad Levels/Radiological Effluent UNUSUAL EVENT ALERT SITE AREA EMERGENCY

(

4.3.U.1-3 Gaseous effluent being released exceeds two times SLC 16.11-6 for>

60 minutes as determined by RP procedure.

4.3.U.1-4 Liquid effluent being released exceeds two times SLC 16.11-1 for> 60 minutes as determined by RP procedure.

Note:

If the monitor reading is sustained for the time period indicated in the EAL AND the required assessments (procedure calculations) cannot be completed within this time period, declaration must be made based on the valid radiation monitor reading.

(Continued) 4.3.A.1-2 A valid indication on radiation monitor EMF-36L of> 5.4E+05 cpm forŽ> 15 minutes or will likely continue for Ž--15 minutes, which indicates that the release may have exceeded the initiating condition and indicates the need to assess the release with procedure SH/I/B/2005/001.

4.3.A.1-3 Gaseous effluent being released exceeds 200 times the level of SLC 16.11-6 for

> 15 minutes as determined by RP procedure.

4.3.A.1-4 Liquid effluent being released exceeds 200 times the level of SLC 16.11-1 for> 15 minutes as determined by RP procedure.

Note:

If the monitor reading is sustained for the time period indicated in the EAL AND the required assessments (procedure calculations) cannot be completed within this time period, declaration must be made based on the valid radiation monitor reading.

(Continued) 4.3.S.1-3 Analysis of field survey results or field survey samples indicates dose consequences greater than 100 mRem TEDE or 500 mRem CDE Adult Thyroid at the site boundary.

Note 1:

These EMF readings are calculated based on average annual meteorology, site boundary dose rate, and design unit vent flow rate.

Calculations by the dose assessment team use actual meteorology, release duration, and unit vent flow rate. Therefore, these EMF readings should not be used if dose assessment team calculations are available.

Note 2:

If dose assessment team calculations cannot be completed in 15 minutes, then valid monitor reading should be used for emergency classification.

END 4.3.G.1-3 Analysis of field survey results or field survey samples indicates dose consequences greater than 1000 mRem TEDE or 5000 mRem CDE Adult Thyroid at the site boundary.

Note 1:

These EMF readings are calculated based on average annual meteorology, site boundary dose rate, and design unit vent flow rate.

Calculations by the dose assessment team use actual meteorology, release duration, and unit vent flow rate. Therefore, these EMF readings should not be used if dose assessment team calculations are available.

Note 2:

If dose assessment team calculations cannot be completed in 15 minutes, then valid monitor reading should be used for emergency classification.

END RP/O/A/500,.I Page 2 of 5 GENERAL EMERGENCY

Enclosure(

Abnormal Rad Levels/Radiological Effluent RP/O/A/5000/uo Page 3 of 5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.U.2 Unexpected Increase in Plant Radiation or Airborne Concentration.

OPERATING MODE:

ALL 4.3.U.2-1 Indication of uncontrolled water level decrease of greater than 6 inches in the reactor refueling cavity with all irradiated fuel assemblies remaining covered by water.

4.3.U.2-2 Uncontrolled water level decrease of greater than 6 inches in the spent fuel pool and fuel transfer canal with all irradiated fuel assemblies remaining covered by water.

4.3.U.2-3 Unplanned valid area EMIF reading, increases by a factor of 1000 over normal levels as shown in Enclosure 4.10.

END 4.3.A.2 Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel.

OPERATING MODE:

ALL 4.3.A.2-1 An unplanned valid trip 11 alarm on any of the following radiation monitors:

Spent Fuel Building Refueling Bridge IEMF-15 2EMF-4 Spent Fuel Pool Ventilation IEMF-42 2EMF-42 Reactor Building Refueling Bridge (applies to Mode 6 and No Mode Only)

IEMF-17 2EMF-2 Containment Noble Gas Monitor (Applies to Mode 6 and No Mode Only)

IEMF-39 2EMF-39 (Continued)

(

I

(

(Continued)

Enclosure(

RP/O/A/5000,.,.

Abnormal Rad Levels/Radiological Effluent Page 4 of 5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.A.2-2 Plant personnel report that water level drop in reactor refueling cavity, spent fuel pool, or fuel transfer canal has or will exceed makeup capacity such that any irradiated fuel will become uncovered.

4.3.A.2-3 NC system wide range level <95% after initiation of NC system make-Lup.

AND Any irradiated fuel assembly not capable of being lowered into spent fuel pool or" reactor vessel.

4.3.A.2-4 Spent Fuel Pool or Fuel Transfer Canal level decrease of >2 feet after initiation of makeup.

AND Any irradiated fuel assembly not capable of being fully lowered into the spent fuel pool racks or transfer canal fuel transfer system basket.

f

(

Enclosure(

RP/O/A/5OOO Abnormal Rad Levels/Radiological Effluent Page 5 of 5 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.3.A.3 Release of Radioactive Material or Increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown.

OPERATING MODE:

ALL 4.3.A.3-1 Valid reading on EMF-12 greater than 15 mR/hr in the Control Room.

4.3.A.3-2 Valid indication of radiation levels greater than 15 mR/hr in the Central Alarm Station (CAS) or Secondary Alarm Station (SAS).

4.3.A.3-3 Valid radiation monitor reading exceeds the levels shown in Enclosure 4,10, END

/"

(

Enclosure (

Loss of Shutdown Functions UNUSUAL EVENT ALERT SITE AREA EMERGENCY RP/O/A/ 5000,,J.

Page 1 of 3 GENERAL EMERGENCY 4.4.A.1 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Trip Was Successful.

OPERATING MODE:

1,2,3 4.4.S.1 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Trip Was NOT Successful.

OPERATING MODE:

1 4.4.A.1-1 The following conditions exist:

4.4.S.1-1 The following conditions exist:

Valid reactor trip signal received or required and automatic reactor trip was not successful.

AND Manual reactor trip from the control room is successful and reactor power is less than 5%

and decreasing.

(Continued)

Valid reactor trip signal received or required and automatic reactor trip was not successful.

AND Manual reactor trip from the control room was not successful in reducing reactor power to less than 5% and decreasing.

END EITHER of the following conditions exist:

"* Core Cooling CSF-RED

"* Heat Sink CSF-RED.

END 4.4.G.1 Failure of the Reactor Protection System to Complete an Automatic Trip and Manual Trip was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core.

OPERATING MODE:

1 4.4.G.1-1 The following conditions exist:

Valid reactor trip signal received or required and automatic reactor trip was not successful.

AND Manual reactor trip from the control room was not successful in reducing reactor power to less than 5% and decreasing.

AND (Continued)

Enclosure(

RP/O/A/500g o

Loss of Shutdown Functions Page 2 of 3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.4.A.2 Inability to Maintain Plant in Cold Shutdown.

OPERATING MODE:

5,6 4.4.A.2-1 Total loss of ND and/or RN and/or KC.

AND One of the following:

"* Inability to maintain reactor coolant temperature below 200'F

"* Uncontrolled reactor coolant temperature rise to

> 180'F.

END 4.4.S.2 Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown.

OPERATING MODE:

1, 2,3,4 4.4.S.2-1 Subcriticality CSF-RED.

4.4.S.2-2 Heat Sink CSF-RED.

4.4.S.3 Loss of Water Level in the Reactor Vessel That Has or Will Uncover Fuel in the Reactor Vessel.

OPERATING MODE:

5,6 4.4.S.3-1 Failure of heat sink causes loss of cold shutdown conditions.

AND Lower range Reactor Vessel Level Indication System (RVLIS) decreasing after initiation of NC system makeup.

4.4.So3-2 Failure of heat sink causes loss of cold shutdown conditions.

AND Reactor Coolant (NC) system mid or wide range level less than 11% and decreasing after initiation of NC system

makeup, (Continued)

I

RP/O/A/500-1 Page 3 of 3 Loss of Shutdown Functions UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.4.S.3-3 Failure of heat sink causes loss of cold shutdown conditions.

AND Either train ultrasonic level indication less than 7.25% and decreasing after initiation of NC system makeup.

END

(

Enclosure

(

Enclosure(..,

Loss of Power UNUSUAL EVENT 4.5.U.I Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes.

OPERATING MODE:

1, 2,3,4 4.5.U.1-1 The following conditions exist:

Loss of offsite power to essential buses ETA and ETB for greater than 15 minutes.

AND Both emergency diesel generators are supplying power to their respective essential busses.

OPERATING MODE:

5, 6, No Mode ALERT 4.5.A.1 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Shutdown Or Refueling Mode.

OPERATING MODE:

5, 6, No Mode 4.5.A.1-1 Loss of all offsite and onsite AC power as indicated by:

Loss of power on essential buses ETA and ETB.

AND Failure to restore power to at least one essential bus within 15 minutes.

(Continued)

SITE AREA EMERGENCY 4.5.S.1 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses.

OPERATING MODE:

1,2,3,4 4.5.S.1-1 Loss of all offsite and onsite AC power as indicated by:

Loss of power on essential buses ETA and ETB.

AND Failure to restore power to at least one essential bus within 15 minutes.

4.5.S.2 Loss of All Vital DC Power.

OPERATING MODE:

1, 2,3,4 (Continued)

(Continued)

(

RP/O/A/5000/001 Page I of 2 GENERAL EMERGENCY 4.5.G.1 Prolonged Loss of All (Offsite and Onsite) AC Power.

OPERATING MODE:

1, 2,3,4 4.5.G.1-1 Prolonged loss of all offsite and onsite AC power as indicated by:

Loss of power on essential buses ETA and ETB for greater than 15 minutes.

AND Standby Shutdown Facility (SSF) fails to supply NC pump seal injection OR CA supply to Steam Generators.

AND At least one of the following conditions exist:

Restoration of at least one essential bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is NOT likely (Continued)

Enclosure(

RP/O/A/5000(

Loss of Power Page 2 of 2 UNUSUAL EVENT 4.5.U.1-2 The following conditions exist:

Loss of offsite power to essential buses ETA and ETB for greater than 15 minutes.

AND One emergency diesel generator is supplying power to its respective essential bus.

4.5.U.2 Unplanned Loss of Required DC Power During Cold Shutdown or Refueling Mode for Greater than 15 Minutes.

OPERATING MODE:

5, 6 4.5.U.2-1 The following conditions exist:

Unplanned loss of both unit related busses: EBA and EBD both <112 VDC, and EBB and EBC both <109 VDC.

AND Failure to restore power to at least one required DC bus within 15 minutes from the time of loss.

END ALERT 4.5.A.2 AC power to essential busses reduced to a single power source for greater than 15 minutes such that an additional single failure could result in station blackout.

OPERATING MODE:

1, 2,3,4 4.5.A.2-1 The following condition exists:

AC power capability has been degraded to one essential bus powered from a single power source for > 15 min. due to the loss of all but one of:

SATA ATC D/G A SITE AREA EMERGENCY 4.5.S.2-1 The following conditions exist:

Unplanned loss of both unit related busses: EBA and EBD both <112 VDC, and EBB and EBC both <109 VDC.

GENERAL EMERGENCY Indication of continuing degradation of core cooling based on Fission Product Barrier monitoring.

END AND Failure to restore power to at least one required DC bus within 15 minutes from the time of loss.

END SATB ATD DIG B.

END

(

Fire/Explosion and Security Events UNUSUAL EVENT 4.6.U.1 Fire Within Protected Area Boundary Not Extinguished Within 15 Minutes of Detection OR Explosion Within the Protected Area Boundary.

OPERATING MODE:

ALL 4.6.U.1-1 Fire in any of the following areas not extinguished within 15 minutes of control room notification or verification of a control room fire alarm.

"* Reactor Building

"* Auxiliary Building

"* Diesel Generator Rooms

"* Control Room

"* RN Purnphouse

"* SSF

"* CAS

"* SAS

"* Doghouses

"* FWST

"* Turbine Buildino

"* Service Building

"* Interim Radwaste Building

"* Equipment Staging Building.

"* Monitor Tank Building ALERT 4.6.A.1 Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown.

OPERATING MODE:

1,2,3,4,5,6 4.6.A.I-1 The following conditions exist:

(Non-security events)

Fire or explosion in any of the following areas:

"* Reactor Building

"* Auxiliary Building

"* Diesel Generator Rooms

"* Control Room

"* RN Pumphouse

"* SSF

"* CAS

"* SAS

"* FWST

"* Doghouses (Applies in Mode 1, 2, 3, 4 only).

AND One of the following:

SITE AREA EMERGENCY 4.6.S.1 Security Event in a Plant Vital Area.

OPERATING MODE:

ALL 4.6.S.1-1 Intrusion into any of the following plant areas by a hostile force:

"* Reactor Building

"* Auxiliary Building

"* Diesel Generator Rooms

"* Control Room

"* RN Pumphouse

"* SSF

"* Doghouses

"* CAS

"* SAS.

4.6.S.1-2 Security confirmed bomb discovered/exploded in a vital area.

4.6.S.1-3 Security confirmed sabotage in a plant vital area.

END Enclosure(

4.6.G.1 Security Event Resulting in Loss Of Ability to Reach and Maintain Cold Shutdown.

OPERATING MODE:

ALL 4.6.G.1-1 Loss of physical control of the control room due to security

event, 4.6.G.1-2 Loss of physical control of the SSF and ASP due to security event.

END

Affected safety system parameter indications show degraded performance (Continued)

RP/O/A/5000 I

Page 1 of 3 GENERAL EMERGENCY I

(Continued)

(

Enclosure Fire/Explosion and Security Events ALERT SITE AREA EMERGENCY

(

4.6.U.1-2 Report by plant personnel of an unanticipated explosion within protected area boundary resulting in visible damage to permanent structure or equipment.

4.6.U.2 Confirmed Security Event Which Indicates a Potential Degradation in the Level of Note:

Safety of the Plant.

OPERATING MODE:

All 4.6.U.2-1 Security confirmed bomb device discovered within plant Protected Area and outside Vital Areas.

4.6.U.2-2 Hostage situation/extortion 4.6.U.2-3 A violent civil disturbance within the owner controlled area.

4.6.U.2-4 A credible terrorist threat as determined by security.

END 4.6.A.2 Plant personnel report visible damage to permanent structures or equipment within the specified area required to establish or maintain safe shutdown within the specifications.

Only one train of a system needs to be affected or damaged in order to satisfy this condition.

Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown.

OPERATING MODE:

No Mode 4.6.A.2-1 The following conditions exist:

(Non-security events)

Fire or explosion in any of the following areas:

"* Spent Fuel Pool

"* Auxiliary Building.

"* RN Pumphouse AND One of the following:

Spent Fuel Pool level and/or temperature show degraded performance (Continued)

RP/O/A/5000u1 Page 2 of 3 GENERAL EMERGENCY UNUSUAL EVENT

Enclosuir RP/O/A/50g 3

Fire/Explosion and Security Events Page 3 of 3 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY Plant personnel report visible damage to permanent str1uctures or equipment supporting spent fuel pool cooling.

4.6.A.3 Security Event in a Plant Protected A rea.

OPERATING MODE:

ALL 4.6.A.3-1 Intrusion into plant Protected Area by a hostile force.

END

(

Enclosure(

Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety ALERT SITE AREA EMERGENCY RP/0/A/5000,uj 1 Page 1 of 4 GENERAL EMERGENCY 4.7.U.1 Natural and Destructive Phenomena Affecting the Protected Area.

OPERATING MODE:

ALL 4.7.U.1-1 Tremor felt and valid alarm on the "strong motion accelerograph".

4.7.U.1-2 Tremor felt and valid alarm on the "Peak shock annunciator",

4.7.U.1-3 Report by plant personnel of tornado striking within protected area boundary.

4.7.U.1-4 Vehicle crash into plant structures or systems within protected area boundary.

4.7.U.1-5 Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.

4.7.A.1 Natural and Destructive Phenomena Affecting the Plant Vital Area.

OPERATING MODE:

ALL 4.7.A.1-1 Valid "OBE Exceeded" Alarm on IAD-4,B/8 4.7.A.1-2 Tornado or high winds:

Tornado striking plant structures within the vital area:

0 0

S 0

Reactor Building Auxiliary Building FWST Diesel Generator Rooms Control Room RN Pumphouse SSF Doghouses CAS SAS.

OR (Continued) sustained winds _Ž 74 mph for

> 15 minutes.

(Continued) 4.7.S.1 Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established.

OPERATING MODE:

ALL 4.7.S.1-1 The following conditions exist:

Control Room evacuation has been initiated per AP/1 (2)/A/5500/017 AND Control of the plant cannot be established from the ASP or the SSF within 15 minutes.

4.7.S.2 Other Conditions Existing Which in the Judgement of the Emergency Coordinator/EOF Director Warrant Declaration of Site Area Emergency.

OPERATING MODE:

ALL 4.7.S.2-1 Other conditions exist which in the Judgement of the Emergency Coordinator/EOF Director indicate actual or likely major failures of plant functions needed for protection of the public.

END 4.7.G.1 Other Conditions Existing Which in the Judgement of the Emergency Coordinator/EOF Director Warrant Declaration of General Emergency.

OPERATING MODE:

ALL 4.7.G.1-1 Other conditions exist which in the Judgement of the Emergency Coordinator/EOF Director indicate:

(1) actual or imminent substantial core degradation with potential for loss of containment OR (2) potential for uncontrolled radionuclide releases. These releases can reasonably be expected to exceed Environmental Protection Agency Protective Action Guideline levels outside the site boundary.

END

(

UNUSUAL EVENT

Enclosure Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety ALERT SITE AREA EMERGENCY

(

4.7.U.2 Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant.

OPERATING MODE:

ALL 4.7.U.2-1 Report or detection of toxic or flammable gases that could enter within the site area boundary in arnounts that can affect safe operation of the

plant, 4.7.U.2-2 Report by Local, County or State Officials for potential evacuation of site personnel based on offsite event.

4.7.U.3 Other Conditions Existing Which in tile j udgement of the Emergency Coordinator/EOF Director Warrant Declaration of an Unusual Event.

OPERATING MODE:

ALL 4.7.U.3-1 Other conditions exist which in the judgement of the Emergency Coordinator/EOF Director indicate a potential degradation of the level of safety of the plant.

4.7.A.1-3 Turbine failure generated missiles, vehicle crashes or other catastrophic events causing visible structural damage on any of the following plant structures:

"* Reactor Building

"* Auxiliary Building

"* FWST

"* Diesel Generator Rooms

"* Control Room

"* RN Pumphouse

"* SSF

"* Doghouses

"* CAS

"* SAS (Continued)

END RP/O/A/5000/uo 1 Page 2 of 4 GENERAL EMERGENCY UNUSUAL EVENT

(

Enclosure(.

RP/O/A/5000QI Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 3 of 4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.7.A.2 Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown.

OPERATING MODE:

ALL 4.7.A.2-1 Report or detection of toxic gases within a Facility Structure in concentrations that will be life threatening to plant personnel.

4.7.A.2-2 Report or detection of flammable gases within a Facility Structure in concentra tions that will affect the safe operation of the plant.

Structures for the above EALs:

"* Reactor Building

"* Auxiliary Building

"* Diesel Generator Rooms

"* Control Room

"* RN Pumphouse

"* SSF

"* CAS

"* SAS (Continued)

/

SEnclosure RP/O/A/5000, Natural Disasters, Hazards, And Other Conditions Affecting Plant Safety Page 4 of 4 UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY 4.7.A.3 Control Room Evacuation Has Been Initiated.

OPERATING MODE:

ALL 4.7.A.3-1 Control Room evacuation has been initiated per AP/I (2)/A/5500/017.

4.7.A.4 Other Conditions Existing Which in the Judgement of the Emergency Coordinator/EOF Director Warrant Declaration of an Alert.

OPERATING MODE:

ALL 4.7.A.4-1 Other conditions exist which in the Judgement of the Emergency Coordinator/EOF Director indicate that plant safety systems may be degraded and that increased monitoring of plant functions is warranted.

END f

.8 RP/O/A/5000/001 Definitions/Acronyms Page I of 3 ALERT-Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA protective action guideline exposure levels.

ALL (As relates to Operating Mode Applicability) - Modes 1,2,3,4,5,6 and No Mode (Defueled)

BOMB-A fused explosive device.

CARF - Containment Air'Return Fan.

CIVIL DISTURBANCE - A group of ten (10) or more people violently protesting station operations or activities at the site. A civil disturbance is considered to be violent when force has been used in an attempt to injure site personnel or damage plant property.

CREDIBLE THREAT - A threat should be considered credible when:

"* Physical evidence supporting the threat exists.

"* Information independent (law enforcement) from the actual threat message exists that supports the threat.

"* A specific group or organization claims responsibility for the threat.

EPA PAG - Environmental Protection Agency Protective Action Guidelines for exposure to a release of radioactive material.

EXPLOSION - A rapid, violent unconfined combustion, or a catastrophic failure of pressurized equipment (e.g., a steamline or feedwater line break) that imparts energy sufficient to potentially damage or creates shrapnel to actually damage permanent structures, systems or components. An electrical breaker flash that creates shrapnel and results in damage to other components beyond scorching should also be considered.

EXTORTION - An attempt to cause an action at the site by threat of force.

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flames is preferred but is NOT required if large quantities of smoke and heat are observed. An electrical breaker flash that creates high temperatures for a short duration and merely localized scorching to that breaker and its compartment should be considered a fire.

FUNCTIONAL - A component is fully capable of meeting its design function. It would be declared INOPERABLE if unable to meet Technical Specifications.

GENERAL EMERGENCY-Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA protective action guideline exposure levels outside the Site Boundary.

.8 RP/O/A/5000/001 Definitions/Acronyms Page 2 of 3 HOSTAGE - A person or object held as leverage against the site to ensure demands will be met by the site.

HOSTILE FORCE - One of more individuals present in a protected area without authorization that may have or have threatened to use force in an attempt to injure site personnel or damage plant property.

IMMINENT - Expected to occur within 1-3 hours.

INOPERABLE - A component does not meet Technical Specifications. The component may be functional, capable of meeting its design.

INABILITY TO DIRECTLY MONITOR - Operational Aid Computer data points are unavailable or gauges/panel indications are not readily available to the operator.

INTRUSION/INTRUDER - Suspected hostile individual present in a protected area without authorization.

LOSS - A component is INOPERABLE and not FUNCTIONAL.

PROLONGED - a duration beyond normal limits, defined as "greater than 15 minutes" or as determined by the judgement of the emergency Coordinator.

PROTECTED AREA - Encompasses all owner controlled areas within the security perimeter fence.

RUPTURED (As relates to Steam Generator) - Existence of primary to secondary leakage of a magnitude sufficient to require or cause a reactor trip and safety injection.

SABOTAGE - Deliberate damage, misalignment, or misoperation of plant equipment with the intent to render the equipment unavailable.

SIGNIFICANT TRANSIENT-An unplanned event involving one or more of the following: (1)

Automatic turbine runback >25% thermal reactor power, (2) Electrical load rejection >25% full electrical load; (3) Reactor Trip, (4) Safety Injection, (5) Thermal power oscillations >10%.

SITE AREA EMERGENCY - Events are in process or have occunred which involve actual or likely major failures of plant functions needed for the protection of the public. Any releases are NOT expected to result in exposure levels which exceed EPA protective action guideline exposure levels outside the Site Boundary.

SITE BOUNDARY - That area, including the protected area, in which Duke Power Company has the authority to control all activities, including exclusion or removal of personnel and property.

SLC - Selected Licensee Commitments.

.8 RP/0/A/5000/001 Definitions/Acronyms Page 3 of 3 SECURITY EVENT - A security related emergency situation for which prompt response by the Security Force, immediate action by plant personnel, and/or assistance from offsite agencies may be required to apprehend intruders and mitigate the effects of or prevent radiological sabotage.

SUSTAINED - A duration of time long enough to confirm that the CSF is valid (not momentary).

TERMINATION - Exiting the emergency condition.

TOTAL EFFECTIVE DOSE EQUIVALENT (TEDE) - The sum of external dose exposure to radioactive plume, to radionuclides deposited on the ground by the plume, and the internal exposure inhaled radionuclides deposited in the body.

TOXIC GAS - A gas that is dangerous to life or health by reason of inhalation or skin contact (e.g.

chlorine).

UNCONTROLLED - Event is not the result of planned actions by the plant staff.

UNPLANNED - An event or action is UNPLANNED if it is not the expected result of normal operations, testing, or maintenance. Events that result in corrective or mitigative actions being taken in accordance with abnormal or emergency procedures are UNPLANNED.

UNUSUAL EVENT-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.

VALID - An indication or report or condition is considered to be VALID when it is conclusively verified by: (I ) an instrument channel check, or (2) indications on related or redundant instrumentation, or (3) by direct observation by plant personnel such that doubt related to the instrument's operability, the condition's existence or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

VIOLENT - Force has been used in an attempt to injure site personnel or damage plant property.

VISIBLE DAMAGE - Damage to equipment or structure that is readily observable without measurements, testing, or analyses. Damage is sufficient to cause concern regarding the continued operability or reliability of affected structure, system, or component. Example damage: deformation due 1

to heat or impact, denting, penetration, rupture, cracking, paint blistering.

VITAL AREA - Areas within the PROTECTED AREA that house equipment important for nuclear safety. Access to a VITAL AREA is allowed only if an individual has been authorized to be in that area.

.9 RP/O/A/5000/001 Emergency Declaration Guidelines Page 1 of 2 THE FOLLOWING GUIDANCE IS TO BE USED BY THE EMERGENCY COORDINATOR IN ASSESSING EMERGENCY CONDITIONS.

"* The Emergency Coordinator shall review all applicable initiating events to ensure proper classification.

The BASIS Document (located in Section D of the Catawba Nuclear Site Emergency Plan) is available for review if any questions arise over proper classification.

Emergencies are declared for the site. If an event results in multiple emergency action levels on a unit or different emergency action levels on each unit, then the emergency declaration shall be based on the higher classification. Information relating to the unit with the lesser classification will be noted as additional information on the Emergency Notification Form (ENF).

If an event occurs, and a lower or higher plant operating mode is reached before the classification can be made, the classification shall be based on the mode that existed at the time the event occurred.

The fission product barrier matrix is applicable only to those events that occur at (Mode 1-4) hot shutdown or higher. An event that is recognized at cold shutdown or lower (Mode 5 or 6) shall not be classified using the fission product banier matrix. Reference would be made to the other enclosures that provide emergency action levels for specific events (e.g. severe weather, fire, security).

"* If a transient event should occur, the following guidance is provided.

i.

Some emergency action levels specify that a condition exist for a specific duration prior to declaration.

a.

For these EALs, the classification is made when the Emergency Coordinator assessment concludes that the specified duration is exceeded or will be exceeded (i.e. condition cannot be reasonably corrected before the duration elapses),

whichever is sooner.

b.

If a plant condition exceeding EAL criteria is corrected before the specified duration time is exceeded, the event is NOT classified by that EAL. Lower Severity EALs, if any, shall be reviewed for possible applicability in these cases.

2.

If a plant condition exceeding EAL criteria is not recognized at the time of occurrence, but is identified well after the condition has occurred (e.g. as a result of routine log or record review) and the condition no longer exists, an emergency shall NOT be declared.

Reporting under 10CFR50.72 may be required. Such a condition could occur, for example, if a follow-up evaluation of an abnormal condition uncovers evidence that the "condition was more severe than earlier believed.

.9 RP/O/A15000/001 Emergency Declaration Guidelines Page 2 of 2

3.

If an emergency classification is warranted, but the plant condition is corrected prior to declaration and notification, the Emergency Coordinator must consider the potential that the initiating condition (e.g. Failure of Reactor Protection System or earthquake) may have caused plant damage that warrants augmenting the on-shift personnel via activation of the Emergency Response Organization. The following action shall be taken:

a.

For UNUSUAL EVENTS, the condition shall be declared and notifications made. The event may be terminated in the same notification or in a follow-up notification.

b.

For ALERT, SITE AREA EMERGENCY, and GENERAL EMERGENCY, the event shall be declared and the emergency response organization activated.

DETERMINATION OF "EVENT TIME" (TIME THE 15 MINUTE CLOCK STARTS)

I.

If plant conditions require implementation of EP/t or 2/A/5000/E-0 (Reactor Trip or Safety Injection), increased emphasis shall be given to evaluation of plant conditions for determination of EAL(s) when "kickout" of the diagnostic procedure occurs. "Event Time" is the time at which the EAL(s) is determined.

2.

If plant conditions do not require implementation of EP/I or 2/A/5000/E-0 (Reactor Trip or Safety Injection), and conditions of a specific EAL are met, the "Event Time" is the time at which the EAL(s) is determined.

3.

The time the event is classified shall be entered on the emergency notification form.

MOMENTARY ENTRY INTO A HIGHER CLASSIFICATION If, while in an emergency classification, the specified EALs of a higher classification are met momentarily, and in the judgment of the Emergency Coordinator are not likely to recur, the entry into the higher classification must be acknowledged. Acknowledgment is performed as follows:

If this condition occurs prior to the initial notification to the emergency response organization and off site agencies, the initial message should note that the site is currently in the lower classification, but had momentarily met the criteria for the higher classification. It should also be noted that plant conditions have improved and stabilized to the point that the criteria for the higher classification are not expected to be repeated.

/

(

Enclosure(

RP/O/A/5004.

Radiation Monitor Readings for Enclosure 4.3 Page 1 of I Note: These values are not intended to apply to anticipated temporary increases due to planned events (e.g. incore detector movement, radwaste container movement, depleted resin transfers, etc.)

Detector Elevation Column Identifier Unusual Event Alert mRad/hr mRad/hr IEMF-1 522' FF, 57 Auxiliary Building Corridor 500 5000 IEMF-3 543' GG, 55 Unit I Charging Pump Area 100 5000 IEMF-4 543' GG, 59 Unit 2 Charging Pump Area 100 5000 IEMF-7 560' NN, 55 Unit I Auxiliary Building Corridor 1500 5000 1EMF-8 560' NN, 59 Unit 2 Auxiliary Building Corridor 500 5000 IEMF-9 577' LL, 55 Unit I Aux. Building Filter Hatch 100 5000 IEMF-10 577' LL, 58 Unit 2 Aux. Building Filter Hatch 100 5000 IEMF-22 594' KK, 53 Containment Purge Filter Area 100 5000 2EMF-9 594' KK, 61 Containment Purge Filter Area 100 5000

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ML020770296

Text

Subject:

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