ML103200310

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Revision 2 to Emergency Plan Implementing Procedure IP-EP-360, Core Damage Assessment.
ML103200310
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 11/04/2010
From: Vogle R
Entergy Corp
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
FOIA/PA-2011-0181, FOIA/PA-2011-0262 IP-EP-360, Rev 2
Download: ML103200310 (22)
v • d • e


Text

. November 4, 2010 DISTRIBUTION CONTROL LIST Document Name: IPEC EMERGENCY PLAN CC# NAME DEPARTMENT LOCATION CC/STMP CROULET, DON INSTRUC TECH TRNG (E-PLAN ONLY) 48-2-A CC/STMP IRAOLA, TONY FOR THE JIC EOF CC/STMP SHIFT MANAGER OPERATIONS IP3 CC/STMP CONTROL ROOM OPERATIONS IP3 CC/STMP EOF E-PLAN (ALL EP'S), EOF CC/STMP PEREZ, ROSE E-PLAN (ALL EP'S) WPO-12D CC/STMP TSC (IP3) EEC BUILDING IP2 CC/STMP BARR, STEVE NRC (ALL EP'S) OFFSITE CC/STMP BARR, STEVE .- __ NRC (ALL EP'S) OFFSITE

-CC/STMP DOC CONTROL DESK NRC (ALL EP'S) OFFSITE CC/STMP DOC CONTROL DESK NRC FOR (E-PLAN ONLY) OFFSITE (USE ATTENTION TO DIRECTOR OF SPENT FUEL ADDRESS)

CC/STMP CULLINAN, P J A (PLAN ONLY) OFFSITE CC/STMP E-PLAN STAFF E-PLAN (ALL EP'S) GSB-2ND FL CC/STMP MCDONALD,MICHELLE ST. EMERG. MGMT. OFFICE (ALL) OFFSITE CC/STMP DELBORGO, D (PLAN ONLY) DISASTER & EMERGENCY OFFSITE CC/STMP LONGO, N (PLAN ONLY) EMERGENCY SERVICES OFFSITE CC/STMP KARSTEN,C (PLAN ONLY) DISASTER & CIVIL DEFENSE OFFSITE CC/STMP STIEBELING A (PLAN ONLY) OFF OF EMERG MANAGEMENT OFFSITE CC/STMP GRANT, LEAH SIMULATOR (TRAINING) 48-2-A CC/STMP GRANT,LEAH LRQ TRAINING 48-2-A CC/STMP CONTROL ROOM OPERATONS IP2 CC/STMP CHIUSANO, J SIMULATOR (TRAINING 5 COPIES) EEC CC/STMP CHIUSANO, J CLASSROOM 2 EEC CC/STMP CHIUSANO, J TRAINING 48-2-A CC/STMP NRC RESIDENT INSPECTOR US NRC (88' ELEVATION) IP2 TONY IRAOLA GETS:.... E-PLAN, IP-EP-115 (FORMS), IP-EP-260(JOINT CENTER INFORMATION) "NO FORMS GO TO THE OFFSITERS"

      • CC/STMP.....CONTROL COPY STAMP***

EFFECTIVE DATE:

IPEC SITE QUALITY RELATED IP-SMM-AD-103 Revision 0 E----ELe°gy 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 1

-Ent CONTROLLED DOCUMENT TRANSMITTAL FORM - PROCEDURES IPEC, P.O. Box 308, Buchanan, NY 10511 TO: DISTRIBUTION DATE: 11/4/10 TRANSMITTAL NO:

,Circle one)

FROM: IPEC DOCUMENT CONTROL: E or IP2 53'EL PHONE NUMBER: (914) 271-7054 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: IPEC EMERGENCY PLAN PROCEDURES DOC# I REV# I TITLE INSTRUCTIONS THE FOLLOWING PROCEDURE HAS BEEN REVISED, PLEASE REMOVE YOUR CURRENT COPY AND REPLACE WITH ATTACHED REVISIED PROCEDURE:

IP-EP-360 REV.2

                      • 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 CU IN V \of- I) 1-4/,vIr L:kr- n ,, -4 )

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O~3*/*I Purtr-UT11LJP- A~r" I,

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CON 'u" FIG LLE0 aE IPEC NON-QUAuTY RELATED IP-EP-360 Revision 2 y EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 1 of 20 CORE DAMAGE ASSESSMENT Prepared by: Robert Vogle Print Name 4VMwx1i,11 /DatVe Approval: Brian A. Sullivan Date Print Name Effective Date: November 4, 2010 This procedureexcluded from further L*-100 reviews.

IP-EP-360 (Core) R2.doc

IPEC NON-QUALITY RELATED IP-EP-360 Revision 2 EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 2 of 20 Table of Contents 1.0 PURPO SE .............................................................................................................................................. 3

2.0 REFERENCES

........................................................................................................................................ 3 3.0 DEFINITIO NS ......................................................................................................................................... 3 4.0 RESPONSIBILITIES ...................................................... 4 5.0 DETAILS .................................................................................................................................................. 4 6.0 INTERFACES ......................................................................................................................................... 8 7.0 RECO RDS ............................................................................................................................................... 8 8.0 REQ UIREM ENTS AND CO MMITM ENTS ......................................................................................... 8 9.0 ATTACHM ENTS ........................................................................................................... I.......................... 8 9.1 Attachm ent 1, Fuel Rod Clad Dam age ........................................................................................... 9 9.2 Attachm ent 2, Fuel Overtem perature Dam age ............................................................................ 14

- nerg IMPLEMENTING PROCEDURES REFERENCE USE Page 3 of 20 CORE DAMAGE ASSESSENT 1.0 PURPOSE 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.

9 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 7000 F, 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 DEFINITIONS None

En/Ier IPEC EMERGENCY PLAN IMPLEMENTING NON-QUAuTY RELATED PROCEDURE I IP-EP-360 Revision 2 PROCEDURES REFERENCE USE. Page 4 of 20 4.0 RESPONSIBILITIES 4.1 Upon recognition of EITHER core exit thermocouple temperature.(s) > 700 'F OR containment radiation levels > 1 R/hr, the Reactor Engineer shall implement this procedure to assess the existence and extent of core damage.

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

5.0 DETAILS 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.

NOTE:

Containment Hi Range Radiation Monitor R-25 and R-26 bottom scale reading is approximately -1 R/hr. Because of this scale limitation of R-25 and R-26, radiation monitors R-2, VC 80ft and R-7, VC Seal table should be used to observe anrincreasing trend towards 1 R/hr (1000 mr/hr), when assessing core damage using the "High level-Core Damage Assessment Flowchart'. Due to containment positions, R-2/R-7 readings of approximately 200 mr/hr, should relate to 1 R/hr on R-25/R-26.

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

P Y I. i IPEC NON-QUAUTY RELATED IP-E *P-360I Revision 2 E*ngyS EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 5 of 20 High Level Core Damage Assessment Flowchart Possible fuel over-.

2f temperature damage, go to Attachment

A IPEC NON-QUALITY RELATED PROCEDURE I IP-EP-360 Revision 2 i nery EMERGENCY PLAN IMPLEMENTING PROCEDURES REFERENCE USE Page 6 of 20 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.OOE+03 2.50E+03 I

2.OOE+03 0)

'0 04 cc 1.OOE+03 5.OOE+02 O.OOE+OO 0 1 2 3 S d4 5 6 7 Time Since Shutdown (hr)

IPEC NON-QUALITY RELATED IP-EP-360 Revision 2 EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 7 of 20 Figure 1B Containment Radiation Level for 1% Fuel Overtemperature Release

(>5 hours after shutdown) 1.40E+03 1.20E+03 1.OOE+03 8.OOE+02 u2) i 6.OOE+02 c\-

4.0OE+02 O.OOE+O0 0 5 10 15 20 25 30 Time Since Shutdown (hr)

a IPEC NON-QUALITY RELATED Enlewo, EMERGENCY PLAN PROCEDURE IPEP360 Revision 2

  • e'* IMPLEMENTING PROCEDURES REFERENCE USE Page 8 of 20 6.0 INTERFACES 6.1 IP-EP-120, Emergency Classification 6.2 IP-EP-220, Technical Support Center 7.0 RECORDS This procedure generates completed Fuel Rod Clad Damage (Attachment 1) and/or Fuel Overtemperature Damage (Attachment 2) worksheets.

8.0 REQUIREMENTS AND COMMITMENTS None 9.0 ATTACHMENTS 9.1 Attachment 1, Fuel Rod Clad Damage 9.2 Attachment 2, Fuel Overtemperature Damage

IP-EP-360 Revision 2 Fnit E

ls IPEC EMERGENCY PLAN IMPLEMENTING NON-QUAUTY RELATED PROCEDURE I PROCEDURES REFERENCE USE Page 9 of 20 Attachment 1 Fuel Rod Clad Damage Sheet 1 of 5 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 2NB) 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:

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

  • Number of CETs at or above 1400°F E=
  • Number of CETs afor abo-v0e1200°F F=

2.2 For RCS pressure at or above 1600 psig:

E x 100

% Clad DamageCET = ------------ =

D 2.3 For RCS pressure below 1600 psig::

F x 100

% Clad Damage CET = --------.. .---

D

r LfEni Enl'gy IPEC EMERGENCY PLAN IMPLEMENTING

} NON-QUAUTY RELATED PROCEDURE t IP-EP-360 Revision 2 PROCEDURES REFERENCE USE Page 10 of 20 Attachment 1 Fuel Rod Clad Damage Sheet 2 of 5 Figure 2A Containment Radiation Level for 100% Clad Damage 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) 1.80E+04 --- RSpressure >160.0 psig, NO containment spray RCS pressure >1600 psig, NO containment spray RCS pressure <1600 psig, NO containment spray RCS pressure <1600 psig, with containment spray 1.40E+04 1.20E+04 T) 1.20E+04 a)

W) 0 8.OOE+03 N

cc) 6.OOE+03 4.OOE+03 2.OOE+03 -

0.OOE+00...........

0 1 2 3 4 5 6 Time Since Shutdown (hr)

Aft IPEC NoN-QuAuTY RELATED IP-EP-360 Revision 2 EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 11 of 20 Attachment 1 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)

I I I I 4.00E+03 ---- RCS pressure >1600 psig, NO containment spray

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

-e-- RCS pressure <1600

>1600 psig, psig, with containment spray r 3.60E+03 3.20E+03 2.80E+03 2.40E+03 \ kN" 0

CID 2.OOE+03 04 cc 1.60E+03 \\-A 1.20E+03 8.OOE-i02-4.OOE+02 0.OOE+00 0 5 1C 0 15 2.0 25 30 Time Since Shutdown (hr)

IPEC NON-QUALITY RELATED IP-EP-360 Revision 1 EMERGENCY PLAN IMPLEMENTING PROCEDURES J PROCEDURE REFERENCE USE Page 12 of 19 Attachment 1 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 (OF)

" Hot leg RTD temperature (OF) 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 Tsat and 650°F?
  • 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

% Diffcdi, --------------------- -------------------------------------- x 100

% 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 0 Bleed paths from the RCS
  • Direct radiation to the containment radiation monitors

IPEC NON-QUALITY RELATED IP-EP-360 Revision 2 QEn1erg EMERGENCY PLAN PROCEDURE I t'* IMPLEMENTING PROCEDURES REFERENCE USE Page 13 of 20 Attachment 1 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 Manager.

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-QUALITY RELATED IP-EP-360 Revision 2 EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 14 of 20 Attachment 2 Fuel Overtemperature Damage Sheet 1 of 7 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 3A/B) G

" Current containment radiation level (R/hr) H 1.3 Estimate fuel overtemperature damage (%):

H x 100

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

G

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

-2.i Determine the following:

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

  • Number of CETs at or above 2000OF K=

2.2 Estimate fuel overtemperature damage (%):

K x 100

% Core Damage cET = --------------

J

IPEC NON-OUALITY RELATED IP-EP-360 Revision 2 EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 15 of 20 Attachment 2 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+O5 2.50E+05 2.OOE+05 0 1.50E+.05 0

LO)

C\1 1.OOE+05 5.OOE+04 O.OOE+O0 0 1 2 3 4 5 6 Time Since Shutdown (hr)

IPEC EMERGENCY PLAN NON-QUALITY PROCEDURERELATED I-P30 Rvso IMPLEMENTING PROCEDURES REFERENCE USE Page 16 of 20 Attachment 2 Fuel Overtemperature Damage Sheet 3 of 7 Figure 3B Containment Radiation Level for 100% Fuel Overtemperature Release

(>5 hours after shutdown)

RCS pressure >1600 psig, NO containment spray 1.40E+05 -*- RCS pressure <1600 psig, NO containment spray

-m- RCS pressure >1600 psig, with containment spray I-4 RCS pressure <1600 psig, with containment spray 1.20E+05 -

1.2OE+05 U) 8.OOE+04 a)

L0 0l cc 6.OOE+04 4.OOE+04 2.OOE+04 -

0.OOE+00 0 5 10 15 20 25 30 Time Since Shutdown (hr)

IPEC NON-QUALITY RELATED IP-EP-360 Revision 2 EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 17 of 20 Attachment 2 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. %)

Below 1050 Yes CH2 No CH3 At or above1050 Yes CH4 No .CH3 3.2 Estimate fuel overtemperature damage (%):

Lx 100

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

M

IPEC NON-QUALITY RELATED IP-EP-360 Revision 2 EMERGENCY. PLAN PROCEDURE E IMPLEMENTING PROCEDURES REFERENCE USE Page 18 of 20 Attachment 2 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.

10 9

8 7

C 0

C ci) 6 0

C 0

0 C

ci) 5 0) 0

  • 0 I

C ci) 4 E

C C

0 3 0

2 1

0 0 5 10 15 20 25 30 35 40 45 50 55 60 Containment Pressure (psig)

IPEC NON-QUALITY RELATED IP-EP-360 Revision 2

- EMERGENCY PLAN PROCEDURE

'* IMPLEMENTING PROCEDURES REFERENCE USE Page 19 of 20 Attachment 2 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 (TF) 4.2 Compare estimated core damage to expected response by answering the following questions (yes/no)

  • 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 cETi

% Diffdiff ---------------------------------------------------------- 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

%/9Diffdiffý.*=-,-. -x 100

% Core Damage Hyd

  • 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

%Diffdiff = ---------------------------------------------------------- x100

% Core Damage Hyd

IPEC NON-QUATY RELATED IP-EP-360 Revision 2 E EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 20 of 20 Attachment 2 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:

0 Injection of water to the RCS

  • Bleed paths from the RCS a 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 Coordinator/TSC Manager.
6. Return to Step 5.1 of the procedure to continue assessment of the reactor core.
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