ML103200310

Revision 2 to Emergency Plan Implementing Procedure IP-EP-360, Core Damage Assessment.

ML103200310
Person / Time
Site:	Indian Point
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)

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***

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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

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IP-EP-360 REV.2

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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

• IPEC. NON-QUALITY RELATED IP-EP-360 Revision 2 EMERGENCY PLAN PROCEDURE

- 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:

• Total number of operable CETs. D=

(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:

• Total number of operable CETs. J=

(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 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 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.