ML040150622

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Revised Emergency Plan Implementing Procedures
ML040150622
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 12/18/2003
From:
Entergy Nuclear Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML040150622 (36)
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Text

12-DEC-03 Page:

16

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~~Entegy MANAGEMENT ADMINISTRATIVE PROCEDURE MANUAL INFORMATIONAL USE Page 13 of 21 ATTACHMENT 10.1 SMM CONTROLLED DOCUMENT TRANSMITTAL FORM SITE MANAGEMENT MANUAL CONTROLLED DOCUMENT TRANSMITTAL FORM - PROCEDURES Page 1 of I

~En tv lergy CONTROLLED DOCUMENT

_________________________________I TRANSMITTAL FORM - PROCEDURES TO: DISTRIBUTION DATE:

12/24/2003 TRANSMITTAL NO: 28902 (Circle one)

FROM: IPEC DOCUMENT CONTROL: EEC or P2 53'EL PHONE NUMBER: 271-7057 The Document(s) identified below are forwarded for use. In accordance with IP-SMM-AD-103, please review to verify receipt, incorporate the document(s) into your controlled document file, properly disposition superseded, void, or inactive document(s). Sign and return the receipt acknowledgement below within fifteen (15) working days.

AFFECTED DOCUMENT:

EMERGENCY PLANNING IMPLEMENTATION PROCEDURES DOC#

REV#

TITLE INSTRUCTIONS NOTE: REPLACE CURRENT INDEX WITH ATTACHED REVISED INDEX.

FOLLO WATTACIED INSTRUCTIONS *

                      • PLEASE NOTE EFFECTIVE DATE***********

RECEIPT OF THE ABOVE LISTED DOCUMENT(S) IS HEREBY ACKNOWLEDGED. I CERTIFY THAT ALL SUPERSEDED, VOID, OR INACTIVE COPIES OF THE ABOVE LISTED DOCUMENT(S) IN MY POSSESSION HAVE BEEN REMOVED FROM USE AND ALL UPDATES HAVE BEEN PERFORMED IN ACCORDANCE WITH EFFECTIVE DATE(S) (IF APPLICABLE) AS SHOWN ON THE DOCUMENT(S).

NAME (PRINT)

SIGNATURE DATE CC#

NAME (PRINT)

SIGNATURE DATE CC#

TO:

ONI:

SUBJECT:

Nuclear Regulatory Commission Controlled Copy #

5 JPEC Emergency Planning Emergency Planning Document Update Date: 12/18/03 Please update your controlled copy of the document listed below as specified with the copy(s) attached.

Docunih Doienfm '

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A Docurneiit#

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New R Old Rev i Emere6 lnIpeetn ae

' -Date

t..yArcin Procedure:.

Emergency Plan Implementing TOC Procedures 12/18/03 12/10/03 Remove and Replace Re.v.0 IP-EP-360 Core Damage Assessment 0Add New Procedure 12/18/03AdNePrcue IP-EP-3 10 Dose Assessment Re3 Remove and Replace pages 9 - 18 12/10/03 Rev. 9 IP-1028 Core Damage Assessment VOID Remove from Binder 06/98 Rev. 31 IP-1070 Inventory VOID Remove from Binder 2/01 Page I of I

Indian Point Energy Center Emergency Plan Implementing Procedures Table of Contents T d I Rezj Effed ctve, IPEC PROCEDURES IP-EP-115 Emergency Plan Forms 6

12/10/03 IP-EP-120 Emergency Classification 0

11/06/03 IP-EP-130 Emergency Notifications and Mobilization 1

12/10/03 IP-EP-212 Unit 2 Control Room 0

12/10/03 IP-EP-213 Unit 3 Control Room 0

12/10/03 IP-EP-222 Unit 2 Technical Support Center 0

12/10/03 IP-EP-223 Unit 3 Technical Support Center 0

12/10/03 IP-EP-232 Unit 2 Operations Support Center 0

12/10/03 IP-EP-233 Unit 3 Operations Support Center 0

12/10103 IP-EP-240 Security 0

12110103 IP-EP-250 Emergency Operations Facility 1

12/10/03 IP-EP-251 Alternate Emergency Operations Facility 2

12/10/03 IP-EP-260 Joint News Center 0

03/06/03 IP-EP-310 Dose Assessment 3

12/10/03 IP-EP-320 Radiological Field Monitoring 0

12/10/03 IP-EP-330 Airborne Sample Analysis 0

12/10103 IP-EP-350 Emergency Contamination Control 0

12/10/03 IP-EP-360 Core Damage Assessment 0

12/18/03 IP-EP-410 Protective Action Recommendations 3

12110103 IP-EP-430 Site Assembly, Accountability & Relocation of Personnel Offsite 1

12/10/03 IP-EP-510 Meteorological, Radiological & Plant Data Acquisition System 2

12/10/03 IP-EP-520 Modular Emergency Assessment & Notification System (MEANS) 2 12/10/03 IP-EP-610 Emergency Termination and Recovery 1

03/06/03 IP-EP-620 Estimating Total Population Exposure

. 1 03/06/03 IP-EP-630 Onsite Medical Emergency 0

11/18/03 6-Pagel1 of 2 As of 12/03

Indian Point Energy Center Emergency Plan Implementing Procedures Table of Contents UNIT 3 PROCEDURES IP-1028 Core Damage Assessment 9

06/98 IP-1052 Hazardous Waste 8

07102 IP-1055 Fire Emergency Response 15 04/02 IP-1057 Natural Phenomena 8

10/01 IP-1059 Air Raid Alert 7

05/01 Inventory IP-1070 Void (Incorporated Into ADS)

IP-2603 Corporate Support Group Manager 1

07/02 Page I-of tL

I e-..Entergy IP-1028 REV.9(VOID DATE 18-DEC-2003)

IS VOID REASON FOR VOID: REPLACED BY IP-EP-360

A--

Entergy IP-1070 REV.31 (VOID DATE 30-DEC-2003)

IS VOID REASON FOR VOID: REPLACED BY IP-EP-AD6

.a IPEC NON-QUALITY RELATED IP-EP-360 Revision 0 Energy.

EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page I

of i

CORE DAMAGE ASSESSMENT CONTROLLED 001 COPY #

Prepared by:

C. Kelly Walker

.C.- VI qI0 3lo Print Name Signature Date Approval:

Frank nzirillo Print Name Date

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5rt I t Effective Date:

IP-EP-360 (Core) RO.doc

af IPEC NON-QUALITY RELATED IP-EP-360 Revision 0 Enkergy EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 2

of 19 Table of Contents 1.0 PURPOSE.........................................

3

2.0 REFERENCES

.3 3.0 DEFINITIONS.3 4.0 RESPONSIBILITIES

.3 5.0 DETAILS

.4 6.0 INTERFACES

.7 7.0 RECORDS

.7 8.0 REQUIREMENTS AND COMMITMENTS

.7 9.0 ATTACHMENTS

.7 9.1, Fuel Rod Clad Damage

.8 9.2 Attachment 2, Fuel Overtemperature Damage.13

isIPEC NON-QUAUTY RELATED IPE-6 Rvso0 I Enegy.

EMERGENCY PLAN PROCEDURE IP-EP-360 Revision 0 e

IMPLEMENTING PROCEDURES REFERENCE USE Page 3

of is CORE DAMAGE ASSESSENT 1.0 PIIRPflSF This guideline provides a methodology for the assessment of:

The degree of damage to the fuel rod cladding that results in the release of the fission product inventory in the fuel rod gap space.

The degree of core overheating that results in the release of the fission product inventory in the fuel pellets.

The appropriate Emergency Action Level for off-site radiological protective actions based on the degree of damage to the reactor core.

This guideline should be used when the reactor is shutdown and either:

Core temperatures are at or above 7000F, or Containment radiation level is at or above 1 R/hr

2.0 REFERENCES

2.1 WCAP-14696-A, Westinghouse Owners Group Core Damage Assessment Guideline, Rev. 1 2.2 "Containment Radiation Level Using Core Damage Assessment Guideline, Revision 1 (1996) For Specific Indian Point Unit 2 EAL Application: A Summary," by Dave Smith, 12/2000.

2.3 PGI-00467-00, 4/5/01 "Containment Radiation Monitor Response/Core Damage Assessment Procedure Support' 2.4 IP-CA-3, Hydrogen Flammability in Containment, Pg 2, Rev. 0 3.0 DEFINITIOlNS None 4.0 BFRPONSIRIL ITIES 4.1 Upon recognition of EITHER core exit thermocouple temperature(s) > 700 F OR containment radiation levels > I R/hr, the Core Physics Engineer (Reactor Engineer) shall implement this procedure to assess the existence and extent of core damage.

4.2 The Core Physics Engineer (Reactor Engineer) shall immediately inform the Technical Assessment Coordinator /TSC Manager of the results of any core damage assessment performed.

is IPEC NON-QUALITY RELATED IP-EP-360 Revision 0

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Page 4

of 1s 5.0 nFTAILS NOTE:

Core Damage Estimate may be base on historical monitor readings. For Example: If core thermocouple readings were high 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into an event but are now off-scale or inoperable use values and time after shutdown for when readings were valid.

5.1 Determine the possible status of the reactor core using the following flowchart and perform the associated action.

High Level Core Damage Assessment Flowchart start

- is IPEC NON-QUALITY RELATED IP-EP-360 Revision 0 Enlerg,'

EMERGENCY PLAN PROCEDURE I

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[_Page 5

of 19 Figure 1A Containment Radiation Level for 1% Fuel Overtemperature Flowchart (0 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after shutdown) 3.00E+03 2.50E+03 2.00E+03 a) a) 1.50E+03 0

CD N

N 1.OOE+03 5.00E+02 O.OOE+00 0

1 2

3 4

5 6

7 Time Since Shutdown (hr)

IPEC NON-QUALITY RELATED Entery EMERGENCYPLAN PROCEDURE 1

IP-EP-360 RevisionO

-i' IMPLEMENTING PROCEDURES REFERENCE USE Page fi of 19 Figure B Containment Radiation Level for 1% Fuel Overtemperature Release

(>5 hours after shutdown) 1.40E+03 1.20E+03 1.OOE+03 6 8.OOE+02 cu a:

to.OOE+02 4.OOE+02 2.00E+02 0.OOE+00

+4-RCS pressure >1600 psig, NO containment spray

._______ -i-RCS pressure <1600 psig, NO containment spray

.- - RCS pressure >1600 psig, with containment spray RCS pressure <1600 psig, with contain ent spray

.~~~~~~~~~~

k11

.~tL~

0 5

10 15 20 25 30 Time Since Shutdown (hr)

IPEC NON-QUALITY RELATED IP-EP-360 Revision 0

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of 12 6.0 INTFRFACFS 6.1 IP-EP-120, Emergency Classification 6.2 IP-EP-222, Unit 2 Technical Support Center 6.3 IP-EP-223, Unit 3 Technical Support Center 7.0 RBEORDS This procedure generates completed Fuel Rod Clad Damage (Attachment 1) and/or Fuel Overtemperature Damage (Attachment 2) worksheets.

8.0 RFO"IIRFMFNT5 ANn C0MMITMFNTS:

None 9.0 ATTArCHMFNTq 9.1, Fuel Rod Clad Damage 9.2, Fuel Overtemperature Damage

i IPEC NON-QUALITY RELATED IP-EP-3W RevIon 0 Ente "y EMERGENCY PLAN PROCEDURE I

IMPLEMENTING PROCEDURES REFERENCE USE Page B

of is Fuel Rod Clad Damage Sheet 1 of 5

1.

Estimate fuel rod clad damage based on containment radiation (CRM) levels.

.1.1 Determine the following:

  • Time since shutdown (hr)

RCS pressure (psig)

Containment sprays operating (yes/no) 1.2 Find the following containment radiation dose rates:

Containment radiation level (R/hr) for 100% clad damage (Figure 2AB)

A= =

Current containment radiation level (R/hr)

B =

1.3 Estimate clad damage ():

B x 100

% Clad Damage CRM = -- ----------

A

2.

Estimate fuel rod clad damage based on Core Exit Thermocouples (CETs).

2.1 Determine the following:

Total number of operable CETs.

D =

(Refer to PICS [Unit 2] or SPDS [Unit 3])

Number of CETs at or above 14000F E =

Number of CETs at or abovel2000F F =

2.2 For RCS pressure at or above 1600 psig:

Ex100

% Clad Damage cET= ------------

D 2.3 For RCS pressure below 1600 psig::

F x 100

% Clad Damage ET = ------------ =

D

IPEC NON-QUALITY RELATED EnWery EMERGENCY PLAN PROCEDURE I

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of IS Fuel Rod Clad Damage Sheet 2 of 5 Figure 2A Containment Radiation Level for 100% Clad Damage Release (O to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after shutdown) 1.80E+04 1.60E+04 1.40E+04 (0

15 cc 0to 0

CD cqJ cm 1.20E+04 1.OOE+04 8.OOE+03 6.OOE+03 4.OOE+03 2.OOE+03 O.OOE+OO 0

1.

2 3

4 5

6 Time Since Shutdown (hr)

IPEC NON-QUALITY RELATED IP-EP-360 Revision 0

- EnjeW EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 1l of 19 Fuel Rod Clad Damage Sheet 3 of 5 Figure 2B Containment Radiation Level for 100% Clad Damage Release

(> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after shutdown) 4.OOE+03 3.60E+03 3.20E+03 aI a,

0 0

CD

,CY CY 8c 2.80E+03 2.40E+03 2.00E+03 1.60E+03

- +RCS pressure >1600 psig, NO containment spray.

.A-RCS pressure <1600 psig, NO containment spray RCS pressure >1600 psig, with containment spray RCS pressure <1600 psig, with containment spra 1.20E+03 8.00E+02 4.OOE+02 O.OOE+00 0

5 10 15 20 25 30 Time Since Shutdown (hr)

IPEC INON-QUALITY RELATED IP-EP-360 Revision 0

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PROCEDURE I

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-I Fuel Rod Clad Damage Sheet 4 of 5

3.

Confirm reasonableness of clad damage estimates.

3.1 Determine the following:

Containment hydrogen concentration (vol. %)

RVILS reading (%)

RCS saturation temperature (F)

Hot leg RTD temperature (F) 3.2 Compare estimated clad damage to expected response by answering the following questions (yes/no)

Is containment hydrogen concentration less than 0.5%?

Is RVLIS between 64% and 47%?

Is hot leg RTD between Tw and 6500F?

Is the absolute difference (% Diff) between estimated containment radiation clad damage and estimated core exit thermocouple clad damage less than 50%?

1% Clad Damage CRM - % Clad damage CETI

% Diff dff= ---------------------------------------------------------- x10O

% Clad Damage CRM 3.3 If all of the answers to the questions in Step 3.2 are YES, the expected response has been obtained; continue at Step 4.

3.4 If any answer to the questions in Step 3.2 is NO, the expected response has not been obtained; determine if the deviation can be explained from either:

3.4.1 Accident progression:

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

MEney IPEC NoN-OUALITY RELATED l

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. Page 12 of 19 Fuel Rod Clad Damage Sheet 5 of 5 3.4.2 Conservatisms in the predictive model:

Fuel burnup Fission product retention in the RCS Fission product removal from containment

4.

Report findings 4.1 If clad damage estimates have increased by more than 1% in the past 30 minutes OR Estimates exceed 2% clad damage Then report possible impact on emergency classification based upon Emergency Action Level thresholds to the Emergency Plant Manager/Plant Operations Manger.

4.2 Report clad damage estimate to the Technical Assessment Coordinator/TSC Manager.

5.

Return to Step 5.1 of the procedure to continue assessment of the reactor core.

IPEC NON-QUAUTY RELATED IP-EP-360 Revision 0

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EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 13 of 19 Fuel Overtemperature Damage Sheet 1 of 7

1.

Estimate Fuel Overtemperature Damage Based on Containment Radiation (CRM)

Levels.

1.1 Determine the following:

Time since shutdown (hr)

RCS pressure (psig)

Containment sprays operating (yes/no) 1.2 Find the following containment radiation dose rates:

Containment radiation level (R/hr) for 100% core overtemperature damage (Figure 3AB) G =

Current containment radiation level (R/hr)

H =

1.3 Estimate fuel overtemperature damage (%):

Hx100

% Core Damage CRM = -- -------

G

2.

Estimate fuel overtemperature damage based on Core Exit Thermocouple (CETs).

2.1 Determine the following:

Total number of operable CETs.

J =

(Refer to PICS [Unit 2] or SPDS [Unit 3])

Number of CETs at or above 20000F K =.

2.2 Estimate fuel overtemperature damage (%):

K x 100

% Core Damage CET = -------

=

J

IPEC NON-QUALITY RELATED I-P30 Rvso

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. REFERENCE USE Page 14 of 19 Fuel Overtemperature Damage Sheet 2 of 7 Figure 3A Containment Radiation Level for 100% Fuel Overtemperature Release (0 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after shutdown) 3.OOE+05

+ RCS pressure >1600 psig, NO containment spray

.i-RCS pressure <1600 psig, NO containment spray A

RCS pressure >1600 psig, with containment spra)

RCS pressure <1600 psig, with containment spraq 2.50E+05 A-2.00E+05 1.50E+05 1.OOE3-05 5.00E+04 0.OOE+00 C,

af)

C'J 1

2 3

4

.5 Time Since Shutdown (hr) 6

IPEC NOUALITYRELATED IP-EP-360 Revision 0

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EMERGENCYPLAN PROCEDURE I

  • En ~~~ IMPLEMENTING PROCEDURES REFERENCE USE Page 15 of is Fuel Overtemperature Damage Sheet 3 of 7 Figure 3B Containment Radiation Level for 100% Fuel Overtemperature Release

(>5 hours after shutdown) 1.40E+05 1.20E+05 1.OOE+05 8.OOE+04 6.00E+04 0,

a, a,

U, 0

AD cc Cl:

cm) 4.00E+04 2.OOE+04 O.OOE+00 0

5 10 15 20 25 Time Since Shutdown (hr) 30

IPEC NON-OUALITY RELATED IP-EP-30 Revision 0

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EMERGENCY PLAN PROCEDURE I

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Page i

of 19 Fuel Overtemperature Damage Sheet 4 of 7

3.

Estimate fuel overtemperature damage based on containment hydrogen (Hyd) concentration.

3.1 Determine the following:

RCS pressure (psig)

Current containment hydrogen concentration (vol. %)

L =

Predicted containment hydrogen concentration at 100% core overtemperature, Table 2 (vol. %)

M =

Table 2 - Core Overtemperature Estimate Based on Containment Hydrogen Concentration RCS Pressure (psig)

Water Injection Predicted Containment into RCS?

Hydrogen Concentration from Figure 4 (vol. /o)

Below 1050 Yes CH2 No CH3 At or abovel 050 Yes CH4 No CH3 3.2 Estimate fuel overtemperature damage (%):

L x 100

% Core Damage Hyd = ------------ =

M

-Enle MIPEC NON-QUALITY RELATED IP-EP-360 Revision 0

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IMPLEMENTING PROCEDURES REFERENCE USE Page 17 of 19 Fuel Overtemperature Damage Sheet 5 of 7 Figure 4 Predicted Containment Hydrogen Concentration for 100% Fuel Overtemperature Note: The wet hydrogen curves are used when superheated conditions inside containment exist or when a manual sample is used.

0 O

a, e

C 0

a 0

tm C

a)

CE C

0 CH2w Y 2-CH dry 9

~~ ~~~~~~~-

C H13 wst CH13 dry

-C-14 wet 7- ___

6___

-~~~~~~~~~~~~~~~1K C

1 dry a

~

~ ~ ~ ~ -

0 5

10 15 20 25 30 35 40 45 50 55 60 Containment Pressure (psig)

IPEC NON-QUALITY RELATED I

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PROCEDURES REFERENCE USE Page la of 19 Fuel Overtemperature Damage Sheet 6 of 7

4.

Confirm reasonableness of fuel overtemperature damage estimates.

4.1 Determine the following:

RVILS reading (%)

Hot leg RTD temperature (F) 4.2 Compare estimated core damage to expected response by answering the following questions (yes/no)

Is RVLIS below 47%?

.Is hot leg RTD at or above 650°F?

  • Is the absolute difference (% Diff) between estimated containment radiation core damage and estimated core exit thermocouple core damage less than 50%?

1% Core Damage CRM - % Core damage cErI

% Diff c = ---------------------------------------------------------- x100

% Core Damage CRM

  • Is the absolute difference (% Diff) between estimated containment hydrogen core damage and estimated radiation core damage less than 25%?

1% Core Damage Hyd - % Core damage CRMI

% Diff dff= ---------------------------------------------------------- X100

% Core Damage Hy

  • Is the absolute difference (% Diff) between estimated containment hydrogen core damage and estimated.

core exit thermocouple core damage less than 25%?

1% Core Damage Hyd - % Core damage cETI

% Diffdi =----------------------------------------------------------

x100

% Core Damage "

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Fuel Overtemperature Damage Sheet 7 of 7 4.3 If all of the answers to the questions in Step 4.2 are YES, the expected response has been obtained; continue at Step 6.

4.4 If any answer to the questions in Step 4.2 is NO, the expected response has not been obtained; determine if the deviation can be explained from either:

4.4.1 Accident progression:

Injection of water to the RCS Bleed paths from the RCS Direct radiation to the containment radiation monitors Hydrogen burn in containment or affects of passive autocatalytic hydrogen recombination (Unit 2) 4.4.2 Conservatisms in the predictive model:

Fuel burnup Fission product retention in the RCS Fission product removal from containment

5.

Report fuel overtemperature estimate to the Technical Assessment CoordinatorlTSC Manager.

6.

Return to Step 5.1 of the procedure to continue assessment of the reactor core.

~Enlergy, IPEC EMERGENCY PLAN IMPLEMENTING PROCEDURES NON-QUALiTy RELATED PROCEDURE REFERENCE USE IP-EP-310 Page Revision 3 9

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Sheet I of 2 Sector Wind From 1

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191° to 213° 3

214° to 235° 4

236° to 2580 5

259° to 2800 6

2810 to 3030 7

3040 to 3250 8

3260 to 3480 9

3490to 100 10 110to330 11 340to 550 12*

560to 780 13*

79° to 1000 14*

101° to 1230 15*

1240to 450 16*

146° to 1680 Distance (Meters)

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5.7 E-6 5.0 E-6 5.3 E-5 5.4 E-5 4.8 E-5 6.4 E-5 6.6 E-5 5.4 E-5 3.2 E-5 3.2 E-5 8.8 E-5 8.3 E-6 1.3 E-5 1.4 E-5 1.9 E-5 1.9 E-5 D

2.1 E-5 1.9 E-5 1.5 E-4 1.6 E-4 1.4 E-4 1.8 E-4 1.9 E-4 1.6 E-4 9.9 E-5 9.9 E-5 2.5 E-4 3.0 E-5 4.3 E-5 4.8 E-5 6.1 E-5 6.2 E-5 E

4.3 3.9 2.7 2.7 2.5 3.1 3.2 2.7 1.8 1.8 4.0 6.0 8.5 9.2 1.2 1.2 E-5 E-5 E-4 E-4 E-4 E-4 E-4 E-4 E-4 E-4 E-4 E-5 E-5 E-5 E-4 E-4 F

1.1 E-4 9.6 E-5 4.9 E4 5.0 E-4 4.7 E-4 5.5 E-4 5.6 E-4 5.0 E-4 3.6 E-4 3.6 E-4 6.7 E-4 1.4 E-4 1.9 E-4 2.0 E-4 2.4 E-4 2.5 E-4 G

2.0 E-4 1.8 E-4 7.1 E-4 7.2 E-4 6.8 E-4 7.9 E-4 8.0 E-4 7.2 E-5 5.4 E-4 5.4 E-4 9.2 E4 2.6 E-4 3.3 E-4 3.5 E-4 4.0 E-4 4.0 E-4

  • Plume for these sectors goes over the water before it touches public or private land. Site boundary in these cases is taken to be the landfall point at the sector center.

IPEC NoN-QUALITY RELATED PROCEDURE IP-EP-310 Revision 3

-I2nleIgy, EMERGENCY PLAN IMPLEMENTING PROCEDURES REFERENCE USE Page 10 of 18.1 Sheet 2 of 2 Site Boundary Xp/Q by Pasquill Stability Category Up Valley Plumes (wind speed <4 mIs) Wind Direction from 102" - 209"(1)

Pasquill Categories A

B C

D E

F I

G 5.2 E-7 1.0 E-6 5.0 E-6 1.9 E-5 3.9 E-5 9.6 E-5 1.8 E-4 Site Boundary XpJQ by Pasquill Stability Category Down Valley Plumes (wind speed <4 mas) Wind Direction from 3400 - 101(2)

Pasquill Categories A

B C

D.

E F

G 3.7 E-6 1.0 E-5 3.2 E-5 9.9 E-5 1.8 E-4 3.6 E-4 5.4 E-4 (1) Plume centerline will always cross the site boundary at sector 2. Therefore, the sector 2 XpIQ values are used.

(2) Plume centerline will cross the site boundary at either sector 8 (Pasquill category A) or sector 10 (for Pasquill category B - G)

Enlergy, IPEC EMERGENCY PLAN IMPLEMENTING PROCEDURES NON-QUALr RELATED PROCEDURE REFERENCE USE IP-EP-310 Page w

Revision 3 11 of 18 a

a.2 Xp/Q Values for other Distances Sheet 1 of I Sector Distance (Meters)

Pasquill Categories 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 6.0 6.5 7.0 7.5 8.0 8.5 9.0 9.5 10.0 1608 2412 3216 4020 4824 5628 6432 7236 8040 8844 9648 10452 11256 12060 12864 13668 14472 15276 16080 A

9.5 E-7 6.3 E-7 5.2 E-7 4.4 E-7 3.6 E-7 3.2 E-7 2.8 E-7 2.6 E-7 2.4 E-7 2.1 E-7 2.0 E-7 1.9 E-7 1.8 E-7 1.7 E-7 1.6 E-7 1.5 E-7 1.5 E-7 1.4 E-7 1.4 E-7 B

4.0 E-6 2.1 E-6 8.3 E-7 5.8 E-7 5.0 E-7 4.2 E-7 3.7 E-7 3.5 E-7 3.2 E-7 3.1 E-7 2.7 E-7 2.5 E-7 2.4 E-7 2.3 E-7 2.2 E-7 2.1 E-7 2.0 E-7 1.9 E-7 1.8 E-7 C

1.5 E-5 1.1 E-5 5.0 E-6 3.5 E-6 2.8 E-6 2.0 E-6 1.6 E-6 1.4 E-6 1.2 E-6 9.9 E-7 8.3 E-7 7.5 E-7 6.7 E-7 6.1 E-7 5.5 E-7 5.0 E-7 4.6 E-7 4.2 E-7 4.0 E-7 D

5.0 E-5 5.4 E-5 1.9 E-5 1.4 E-5 1.0 E-5 8.1 E-6 6.8 E-6 5.8 E-6 5.1 E-6 4.4 E-6 3.8 E-6 3.5 E-6 3.2 E-6 3.0 E-6 2.7 E-6 2.5 E-6 2.3 E-6 2.1 E-6.

2.1 E-6 E

9.0 E-5 5.4 E-5 3.9 E-5 3.7 E-5 2.2 E-5 1.8 E-5 1.5 E-5 1.3 E-5 1.1 E-5

  • 1.0 E-5 9.1 E-6 8.2 E-6 7.5 E-6 6.9 E-6 6.3 E-6
  • 5.8 E-6 5.5 E-6 5.4 E-6 5.3 E-6 F

2.1 E-4 1.3 E-4 9.6 E-5 7.0 E-5 5.7 E-5 4.7 E-5 4.0 E-5 3.5 E-5 3.1 E-5 2.8 E-5 2.5 E-5 2.3 E-5 2.1 E-5 1.9 E-6 1.8 E-5 1.7 E-5 1.6 E-5 1.5 E-5 1.5 E-5 G

3.4 E-4 2.2 E-4 1.8 E-4 1.7 E-4 1.3 E-4 1.1 E-4 9.4 E-5 7.3 E-5 6.7 E-5 5.9 E-5 5.A E-5 5.0 E-5 4.7 E-5 4.3 E-5 4.1 E-5 3.8 E-5 3.6 E-5 3.4 E-5 3.4 E-5.

W^eft IPEC SITE NON-QUALITY RELATED IP-EP-310 Revision 3

=EnIegy, EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURE REFERENCE USE Page 12 of 18.3 2, 5 and 10-Mile Xp1Q Values Sheet I of Xii/Q PASQUILL CATEGORY A

B C

D E

F G

2 MILE 5.2E-7 8.3 E-7 5.OE-6 1.9E-5 3.9E-5 9.6E-5 1.8E-4 5 MILE 2.4E-7 3.2E-7 1.2E-6 5.1 E-6 1.1 E-5 3.1 E-5 6.7E-5 10 MILE 1.4E-7 1.8E-7 4.OE-7 2.1 E-6 5.3E-6 1.5E-5 3.4E-5

IPEC SITE NON-QUALITY RELATED IP-EP-310 Revision 3

- Enlergy, EMERGENCY PLAN PROCEDURE IMPLEMENTING a

PROCEDURE REFERENCE USE Page 13 of 18.4 Reuter-Stokes Location XpIQ Values Sheet I of 1 Stability Class Sector Monitor A

B C

D E

F G

Distance (m) 1 3226 5.3E-7 8.4E-7 5.1E-6 1.9E-5 4.OE-5 9.8E-5 1.8E-4 2

3379 5.2E-7 8.3E-7 5.OE-6 1.8E-5 3.9E-5 9.7E-5 1.7E-4 3

2574 6.3E-7 1.2E-6 7.3E-6 2.6E-5 5.3E-5 1.2E-4 2.4E-4 4

1448 1.2E-6 4.6E-6 1.8E-5 6.1E-5 1.1E-4 2.4E-4 3.9E4 5

1287 1.4E-6 6.4E-6 2.3E-5 7.3E-5 1.4E-4 2.8E-4 4.4E-4 6

643 4.3E-6 2.2E-5 6.OE-5 1.8E-4 3.OE-4 5.5E-4 7.7E4 7

643 4.3E-6 2.2E-5 6.0E-5 1.8E-4 3.OE-4 5.5E4 7.7E4 8

804 2.9E-6 1.7E-5 4.5E-5 1.3E-4 2.4E-4 4.5E-4 6.6E-4 9

1126 1.8E-6 8.5E-6 2.6E-5 8.1E-5 1.5E-4 3.2E-4 4.9E-4 10 1287 1.4E-6 6.4E-6 2.3E-5 7.3E-5 1.4E-4 2.8E-4 4.4E-4 11 1287 1.4E-6 6.4E-6 2.3E-5 7.3E-5 1.4E-4 2.8E-4 4.4E-4 12 2494 6.4E-7 1.3E-6 7.5E-6 2.7E-5 5.6E-5 1.2E-4 2.4E-4 13 1870 8.OE-7 2.7E-6 1.2E-5 4.2E-5 8.1E-5 1.8E4 3.2E-4 14 1870 8.0E-7 2.7E-6 1.2E-5 4.2E-5 8.1E-5 1.8E-4 3.2E-4 15 1648 9.4E-7 3.9E-6 1.5E-5 5.OE-5 9.7E-5 2.1E4 3.6E4 16 1770 8.4E-7 3.3E-6 1.3E-5 4.5E-5 8.8E-5 1.9E4 3.4E4

_j Mentergy.,

IPEC SITE NON-QUALITY RELATED IP-EP-310 Revision 3

~ Enlergy, EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURE REFERENCE USE Page 14 of 18.5 Accident Monitoring of Noble Gas Concentrations in the Plant Vent Sheet I of 2 NOTES

1. The Operations Support Center (OSC) H.P Team Leader/ Rad. Protection Coordinator will determine which reading to obtain first; plant vent or back-up plant vent monitoring.
2. Locations and equipment may be different from Unit 2 to Unit 3 1.0 Radiation readings may be obtained on the plant vent by the following:

1.1 Follow the provisions used by the OSC to plan and track team assignments.

1.2 Use a telescoping radiation monitoring instrument (e.g. teletector or equivalent) to perform this function.

1.3 AAs requested by OSC Health Physics (HP) Team Leader or Control Room (CR), REPORT radiation levels.

1.4 Proceed to the Containment Airlock area.

1.5 Using the fan-building wall for shielding, obtain radiation readings by Vapor Containment purge and exhaust ducts.

CAUTION The door leading out to the plant vent area may lock when closed. To prevent being trapped in the plant vent area, BLOCK OPEN THE DOOR prior to going to the plant vent area.

1.6 Proceed through the door to the plant vent area.

1.7 Obtain radiation readings at the following locations:

1.7.1 6 feet from the plant vent 10 feet above the floor.

1.7.2 Contact with the plant vent 10 feet above the floor.

1.8 Notify the OSC or CR that radiation readings have been obtained and follow instructions as directed.

IPEC SITE NON-QUALITY RELATED IP-EP-310 Revision 3

Enlergy, EMERGENCY PLAN PROCEDURE IMPLEMENTiNG PROCEDURE REFERENCE USE Page 15 of 18.5 Accident Monitoring of Noble Gas Concentrations in the Plant Vent Sheet 2 of 2 2.0 Backup plant vent monitoring readings may be obtained by the following:

2.1 Follow the provisions used by the OSC to plan and track team assignments.

2.2 Proceed to the Auxiliary Building (PAB) Post Accident (PASS) Plant Vent Sample Cave 2.3 Ensure that the RMS-2 meter is positioned on top of the PASS plant vent shield.

2.4 Ensure that the RMS-2 detector is positioned on the floor of the PASS plant vent shield near the gas-sampling bulb.

2.5 Ensure that detector is connected properly to meter with the cable run through the 1-inch hole in the top of the PASS plant vent shield.

2.6 Ensure that the meter is energized by A/C and the power is on.

2.7 With the shield door closed, Establish recirculation flow of plant vent gases through the Pass plant vent piping according to RE-CS-040.

2.8 After recirculation is equilibrated (about 5 minutes) 2.9 Record backup plant vent readings from the RMS-2 monitor.

2.10 Using a hand held meter, OBTAIN a background radiation reading outside of the PASS plant vent shield.

2.11 Report RMS-2 readings to the OSC or CR and FOLLOW instructions as directed.

Enlerg IPEC SITE NON-QUALITY RELATED IP-EP-310 Revision 3 B Enlergy EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURE REFERENCE USE Page 16 of 18.6 Manual Calculation of Thyroid CDE Sheet of 1 Thyroid Committed Dose Equivalent (CDE) for Entergy personnel using actual or estimated data for radioiodine concentrations and stay times.

Calculation:

The following Dose Conversion Factors should be used to determine thyroid CDE based on airborne radioiodine concentration:

MIX' DCF = 4.00E08 mRem/hr jCi/cc 1-131 DCF= 1.30EO9 mRem/hr pCi/cc 1-132 DCF = 7.50E06 mRem/hr pCi/cc 1-133 DCF = 2.20E08 mRem/hr pCi/cc 1-134 DCF = 1.30E06 mRem/hr pCi/cc 1-135 DCF = 3.80E07 mRem/hr pCi/cc

'To be used for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown when the radioiodine mix is not known.

The 1-131 DCF is to be used for times greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown when the radioiodine mix is not known.

IF thyroid CDE is expected to exceed 5 Rem for any Entergy personnel, THEN the ORM should RECOMMEND to the ED that KI be issued to these individuals.

___=W IPEC SITE NON-QUALITY RELATED IP-EP-310 Revision 3

Enlergy, YEMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURE REFERENCE USE Page 17 of 18.7 Discussion Sheet 1 of 2 The following instrumentation/methodology can be used to determine the noble gas release rate.

Plant vent monitor-low range (Direct Readout)

Plant vent monitor-high range (Direct Readout)'

Plant vent survey-hand held instrument or remote readout Isotopic analysis of sample taken from release point.

Condenser air ejector monitor (Direct Readout).

Main steam line monitors.

Back calculating a release rate based on actual field radiological data.

Containment radiation monitors R-25 and R-26 to measure the source term within containment and to estimate potential releases from containment.

Potential exposure to the population if a future release of the existing containment source term occurs, utilizing the following information:

1. Containment pressure relief line contains three isolation valves (one in containment and two outside).
2. Containment purge system contains two isolation valves on the Inlet Duct (one in containment and one outside).
3. Containment purge system contains two isolation valves on the Exhaust Duct (one in containment and one outside).
4. Weld Channel (WC) and Isolation Valve Seal Water System (IVSWS) are pressurized to ensure that during accident conditions a pressure build up to AT LEAST 50 psi in containment would NOT cause a leak of radioactive material to the environment as long as the isolation valves remained in the closed position.

En0=ff IPEC SITE NON-QUALITY RELATED IP-EP-310 Revision 3

- E nlergy, EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURE REFERENCE USE Page 18 of 18.6 Discussion Sheet 2 of 2

5. WITHOUT WC AND IVSWS, BUT with isolation valves closed, the containment leak rate is expected to be LESS THAN 0.1% of the containment volume per day (Tech Spec) WITH a pressure buildup to 50 psi inside containment. At lower pressures the leak rate would be smaller, approaching zero as the pressure differential approaches zero.
6. Containment Volume = 2.6 x 10o6 ft =7.4x cc
7. For P2 and Post-Steam Generator Tube Rupture (SGTR) cooldown using blowdown'situations, the determination of the gaseous release rate from the blowdown flash tank shall be accomplished by determining the noble gas concentration in the faulted SG blowdown (Chem sample pCi/cc) AND the blowdown rate (GPM).
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