ML20157A189

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IP-EP-360, Revision 6, Core Damage Assessment
ML20157A189
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 05/21/2020
From: Martin R
Entergy Nuclear Operations
To:
Office of Nuclear Reactor Regulation
References
IP-EP-360, Rev 6
Download: ML20157A189 (28)
v • d • e


Text

--- Entergy Indian Point Energy Center CONTROLLED DOCUMENT Document Control TRANSMITTAL 450 Broadway Buchanan, NY 10511 TO: DISTRIBUTION DATE: 6/1/2020 FROM: IPEC DOCUMENT CONTROL -- 3RD FLOOR ADMIN BLDG PHONE#:

TRANS MITTAL #: EP-20-0022 The Document(s) identified below are forwarded for use. Please review to verify receipt, incorporate the document(s) into your controlled document file, properly disposition superseded, voided, or inactive document(s).

AFFECTED DOCUMENT IPEC EMERGENCY PLANNING PROCEDURES DOC# I REV# I TITLE I INSTRUCTIONS THE FOLLOWING PROCEDURE(S) HAS BEEN REVISED, PLEASE REMOVE YOUR CURRENT COPY AND REPLACE WITH ATTACHED UPDATED REVISION:

IP-EP-360 REVISION 6 EFFECTIVE DATE: 6/1/2020 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 I CERTIFY THAT ALL SUPERSEDED, VOID, OR INACTIVE COPIES OF THE ABOVE LISTED DOCUMENT(S) IN MY POSSESSION HAVE BEEN REMOVED FROM USE AND ALL UPDATES HAVE BEEN PERFORMED IN ACCORDANCE WITH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).

U.S. NUCLEAR REGULATORY COMMISSION ATTN: DOCUMENT CONTROL DESK 11555 ROCKVILLE PIKE ROCKVILLE, MD 20852 NAME (PRINT) SIGNATURE DATE COPY LOCATION

Attachment 1 Page 1 of 2 10CFR50.54(Q 1(2) Review Procedure/Document Number: IP-EP-360 Revision: 6 Equipment/Facility/Other: Indian Point Energy Center

Title:

Core Damage Assessment Part I. Description of Activity Being Reviewed (event or action, or series of actions that have the potential to affect the emergency plan or have the potential to affect the implementation of the emergency plan):

Procedure was revised, to reflect the requirement in the Post Unit 2 Shutdown Eplan (PSEP), as submitted to the NRC per LAR, license #NL-19-001. See attached matrix for*

changes made. Procedure will be effective on June 1, 2020, Part II. Emergency Plan Sections Reviewed (List all emergency plan sections that were reviewed for this activity by number and title. IF THE ACTIVITY IN ITS ENTIRETY IS AN EMERGENCY PLAN CHANGE, EAL CHANGE OR EAL BASIS CHANGE, ENTER THE SCREENING PROCESS. NO 10CFR50.54(q)(2) DOCUMENTATION IS REQUIRED.

Part 1

Introduction:

Section A: Purpose Part 2 Planning Standards and Criteria:

Section A: Assignment of Responsibility Section B: Station Emergency Response Organization Section H: Emergency Facilities and Equipment Section I: Accident Assessment..

Part Ill. Ability to Maintain the Emergency Plan (Answer the following questions related to impact on the

1. Do any elements of the activity change information contained in the emergency plan (Section 3.0 Step 6}?

YES O NO 181 IF YES, enter screening process for that element

2. Do any elements of the activity change an emergency classification Initiating Condition, Emergency Action Level (EAL), associated EAL note or associated EAL basis information or their.underlying calculations or assumptions?

YES O NO 181 IF YES, enter screening process for that element

3. . Do any elements of the activity change the process or capability for alerting and notifying the public as described in the FEMA-approved Alert and Notification System design report?

YES O

  • NO 181 IF YES, enter screening process for that element
4. Do any elements of the activity change the Evacuation Time Estimate results or documentation?

YES O

  • NO 181 IF YES, enter screening process for that element
5. Do any elements of tl:ie activity change the Onshift Staffing Analysis results or documentation?

YES O NO 181 IF YES, enter screening process for that element EN-EP-305 ROOS

Attachment 1 Page 2 of 2 10CFR50.54(Q !(2) Review Procedure/Document Number:* IP-EP-360 Revision: 6 Equipment/Facility/Other: Indian Point Energy Center *

Title:

Core Damage Assessment Part IV. Maintaining the Emergency Plan Conclusion The questions in Part Ill do not representthe sum total of all conditions that may cause a change to or impact the ability to maintain the emergency plan. Originator and reviewer signatures in Part V document that a review of all elements of the proposed change have been considered for.

their impact on the ability to maintain the emergency plan and their potential to change the emergency plan.

1. Provide a brief conclusion that describes how the conditions as described in the emergency plan are maintained
  • with this activity. *
2. Check the box below when the 10CFR50.54(q)(2) review completes all actic;ms for all elements of the activity- no 10CFR50.54(tl)(3) screening or evaluation is required for any element; Otherwisf!, leave the checkbox blank. .

[811 have completed a review of this activity in accordance with 10CFR50.54{q)(2) and determined that the effectiveness of the emergency plan is maintained. This activity does not make any changes to ttie emergency plan. No further actions are required to screen or evaluate this activity under 10CFR50.54(q)(3).

Per Post Shutdown Emergency Plan (PSEP), Unit 3 CCR will be the active/running plant and Unit 2 will be at shut down. The changes made to this procedure (see attached matrix) reflects this requirement of the Post Unit 2 Shutdown Eplan, as submitted to the NijC (license# NL-19-001) and added a graph to support determining RVLIS level for Unit 3. The NRC has approved the PSEP per RA-20-040.

A review ofthis activity in accordance with 10 CFR 50.54(q)(2} has been completed and determined that the effectiveness of the PSEP is maintained. This revision.aligns the procedure with the protocols of the post Unit 2 shutdown. None of the changes afl'.ect the ability to perform classifications, notifications, or PARs, it does not affect activation or staffing of the ERO, and all planning standard requirements are maintained. The changes made do not require a change to the Emergency Action Level scheme, On-shift Staffing study or the PSEP.

No further actions are required to screen or evaluate this activity under 10 CFR 50.54(q)(3).

Part V. Signatures:

Preparer Name (Print) Date:

Rebecca A. Martin 5/21/2020 (Optional) Reviewer Name (Print) Reviewer Signature Date:

Reviewer Name (Print) Reviewer Signature Date:

Timothy Garvey Nuclear EP Project Manager

~elxee"'-a. M~ ,&,, 1:Qt>Jl,JU-\.

Approval Per Telecom Approver Name (Print) Approver Signature Date:

Frank Mitchell Emergency Planning Manager or deslgnee ##I~

EN-EP-305 ROOS

IP-EP-360 Revision 6 REVISION MATRIX Change Page/Section Previous Version New Version Editorial Effect on 10 CFR 50.47(b)

No. Change Planning Standards or NUREG-0654 program elements? Justify if NO.

1. Page 3 Reference 2.2 "Containment Radiation None N N - Removed reference for Unit Level Using Core 2 and updated numbering. Per Damage Assessment Post Shutdown Emergency Plan, Guideline, Revision 1 Unit 3 CCR will be the (1996) For Specific active/running plant and Unit 2 Indian Point Unit 2 EAL will be defueled and core Application: A Summary, damage assessment is not by Dave Smith, 12/2000. , needed for Unit 2. This change reflects that requirement in the Post Unit 2 shut down Eplan,

. which is under an LAR. (license #

NL-19-001) NRC approved per RA-20-040.

2. Page 4 Section 5.1 b Use H2-02 analyzer on
b. Initiate performance of 3-SOP- N N - removed Use H2-02 analyzer
b. SS-004, "containment Hydrogen on Accident Assessment Panel Accident Assessment Panel (Unit Concentration Measurement System" (Unit 2) Per Post Shutdown
2) or Initiate performance of 3-(Unit 3). Emergency Plan, Unit 3 C9R will SOP'-SS-004, "containment be the. active/running plant and Hydrogen Concentration Unit 2 will be defueled and core Measurement System* (Unit 3).

damage assessment is not needed for Unit 2. This change reflects that requirement in the Post Unit 2 shut down Eplan, which is under an LAR. (license #

NL-19-001) NRC approved per RA-20-040.

3.  : Page 5 Section 9.4 9.4 Attachment 2, RCS-15 y N - added an attachment to the None procedure. This change only updated the attachment section.

See Change #8 for justification.

Page 1 of 3

IP-EP-360 Revision 6 REVISION MATRIX

4. Page 9 Section 2.1 (Refer to PICS or SPDS [Unit 31) N N - removed Unit 2 reference.

(Refer to PICS [Unit 2] or SPDS Per Post Shutdown Emergency

[Unit 31)

Plan, Unit 3 CCR will be the

/

active/running plant and Unit 2

  • will be defueled and core damage assessment is not needed for Unit 2. This change reflects that requirement in the Post Unit 2 shut down Eplan, which is under an LAR. (license #

NL-19-001) NRC approved per RA-20-040.

5. Page 12, section 3.1 3.1 Determine the 3.1 Determine the following: N N - added "(see Attachment 4)"

following: (see Attachment 4) to this step to support in determining RVLIS indications.

See justification in change #8. '"

6. Page 14 Section 2.1 (Refer to PICS [Unit 31) N N - Removed Unit 2 reference.

(Refer to PICS [Unit 2] or [Unit 31)

Per Post Shutdown Emergency Plan, Unit 3 CCR will be the active/running plant and Unit 2 will be defueled and core damage assessment is not needed for Unit 2. This change reflects that requirement in the Post Unit 2 shut down Eplan, which is under ~n LAR. (license #

NL-19-001) NRC approved per

' RA-20-040.

7. Page 20, Section 4.4.1

. - Hydrogen burn in containment N N .:.. Removed Unit 2 reference.

Per Post Shutdown Emergency containment or affects of passive Plan, Unit 3 CCR will be the autocatalytic hydrogen active/running plant and Unit 2 recomb_ination (Unit 2) will be defueled and core damage assessment is not needed for Unit 2. This change reflects that requirement in the Post Unit 2 shut down Eplan, which is under an LAR. (license #

- NL-19-001) NRC approved per RA:-20-040.

Page 2 of3

IP-EP-360 Revision 6 REVISION MATRIX

8. Page 21 Attachment 4 RCS-15, Rev. O N N -. Per request o(Rx**
  • None RVLIS F'ull Runge Level lndicut.i.on Map

,oo - , - - - = - ~ - - - , 75 Engineering who is the end user 90 of this procedure, Attachment 4 70

  • * . .B . . . ..

RCS-1-5 graph was added to 70 assistant in determining RVLIS indications. This graph is used by

§:

~ Rx Engineering during drills and CORE 00

~ finding the graph has been time

,o consuming. Graph was added to

. ,s procedure to reduce time spent 40 looking for it. Adding the graph

  • _i:.____~__::,,,.._,=.../__,.,. did not change the intent of the

~~".unn*

.....,,~*==*

lOO~ T,:,p of Vc:o*c,I v.,*** , n .. ,...

procedure and will aide in getting lulwl/C>ulllO"l. H<>UI"*

'l'up c,.f c.,..., (IJCP)

Uoll<nn of c.,,.. Cl.CJ*)

results faster and does not need DoU.ott1 nl \'.,..,..,,

any further evaluation Page 3 of 3

IP-SMM-AD-102 Rev: 17 IPEC IMPLEMENTING PROCEDURE Page 35 of 43 PREPARATION, REVIEW, AND APPROVAL ATTACHMENT 10.2 IPEC PROCEDURE REVIEW AND APPROVAL (Page 1 of 1)

Procedure

Title:

Core Damage Assessment Procedure No* IP-EP-360 Existing Rev* 5 New Rev* 6 ORN/EC No* DRN-20-00308 Procedure Activiti Tempora[Y Procedure Change (MARK Applicable} Converted To IPEC, Replaces: (MARK Applicable}

NEW PROCEDURE Unit 1 Procedure No: EDITORIAL Temporary Procedure Change D GENERAL REVISION

~ PARTIAL REVISION Unit 2 Procedure No: 0 ADVANCE Temporary Procedure Change D EDITORIAL REVISION CONDITIONAL Temporary Procedure Change D VOID PROCEDURE Terminating Condition:

Unit 3 Procedure No:

SUPERSEDED D RAPID REVISION Document in Microsoft Word:

D Yes D No VOID DRNffPC No(s):

Revision Summary D N/A - See Revision Summary Matrix Implementation Requirements Implementation Plan? Yes IBl No Formal Training? 00 Yes No Special Handling? D Yes IBl N?J RPO Dept: Emergency Planning Writer (Print Name/ Sign): Rebecca A. Martin/x7106/ ~elJ;:cr"- ~

Review and Approval (Per Attachment 10.1, IPEC Review n

1. fBJ Technical Reviewer: Kevin Robinson / -S-- {"/ 171-i)

Cross-D~cipElinary Reviewers: /_ //// . Olgitallysign*dbyRaymondW!llams Dept ~~ J\~ Reviewer: _Ra....,Y'-Wi_i_lli_am_s_ ___,,,,,...,.--,-,--7_'~~...,....,~----,-:-o_a*e:..,..2_02_0.o_s.,_60_1,_s2:_*29_.--0_4*00_*_ _ _ __

cJ (Print Name/ Signature/ Date)

Dept: _ _ _ _ _ _ Reviewer:--------,,,--,.-.-,-----,~----=----,.----------

Print Name/ Signature/ Date

3. IE! RPO* Responsibilities/Checklist: F, Mitchell_.._=._,,,,,....,.:~...:...,.:;..;~:..=~~=-.,...:5'f"f.L.~/.,...:5:..;~~~-.;l.e> _ _ _ _ _ __

(Print Name/ Signature/ Date)

IBl PAD required and is complete (PAD Approver and Reviewer qualifications have been verified)

Previous exclusion from further Ll-100 Review is still valid D PAD not required due to type of change as defined in 4.6 4, D Non-Intent Determination Complete: --------=--,----,----------------

(Print Name/ Signature/ Date)

NO change of purpose or scope NO change to less restrictive acceptance criteria NO reduction in the level of nuclear safety NO change to steps previously identified as commitment steps NO voiding or canceling of a procedure, unless NO deviation from the Quality Assurance Program Manual requirements are incorporated into another procedure NO change that may result in deviations from Technical or the need for the procedure was eliminated via an Specifications, FSAR, plant design requirements or previously alternate process. made commitments.

5. D On-Shift Shift Manager/CRS:

(Print Name/ Signature/ Date)

6. User Validation: User:
7. Special Handling Requirements Understood:

~ IPEC NON-QUALITY RELATED Revision 6 IP-EP-360

~Er1tergy.,, EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 1 of 21 CONTROLLED CORE DAMAGE ASSESSMENT i Prepared by: Rebecca A. Martin ~~crv-0-Mr~ ,5k;li) r)O~

I Print Name Signature Date Approval: Frank J. Mitchell Print Name U11!ti£MJSignature s--LM,/2/);o Date Effective Date: June 1, 2020

  • . .) IP-EP-360 (Core) R6.doc

a IPEC NON-QUALITY RELATED IP-EP-360 Revision 6

~ Eitler.oy,.

. . b"'

EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page _g of 21 Table of Contents 1.0 PURPOSE .............................................................................................................................................. 3

2.0 REFERENCES

....................................................................................................................................... 3 3.0 DEFINITIONS ......................................................................................................................................... 3 4.0 RESPONSIBILITIES ............................................................................................................................... 4 5.0 DETAILS .................................................................................................................................................. 4 6.0 INTERFACES ......................................................................................................................................... 5 7.0 RECORDS ............................................................................................................................................... 5 8.0 REQUIREMENTS AND COMMITMENTS ............................................................................................... 5 i) 9.0 ATTACHMENTS ..................................................................................................................................... 5 9.1 Attachment 1,High Level Core Damage Assessment Flowchart ........................................................ 6 Figure 1A, Containment Radiation Level for 1%. 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) .......................................................................................................................... 7 Figure 1 B,Containment Radiation Level for 1%. Fuel Overtemperature Release (>5 hours after shutdown) .......................................................................................................................... 8 9.2 Attachment 2, Fuel Rod Clad Damage ................................................................................................ 9 9.3 Attachment 3, Fuel Overtemperature Damage ................................................................................ 14 9.4 Attachment 4, RCS-15, RVLS Full Range Level Indication Map .............................................. 21

IPEC NON-QUALITY RELATED IP-EP-360 Revision 6 EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page J of 21 CORE DAMAGE ASSESSENT 1.0 PURPOSE This procedure provides a methodology for the assessment of:

e The degree of damage to the fuel rod cladding that resulti:; in the release of the fission product inventory in the fuel rod gap space.

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

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

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

e Core temperatures are at or above 700°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 PGl-00467-00, 4/5/01 "Containment Radiation Monitor Response/Core Damage Assessment Procedure Support" 2.3 IP-CA-3, Hydrogen Flammability in Containment, Pg 2, Rev. 0 3.0 DEFINITIONS None

IPEC NON-QUALITY RELATED Revision 6 IP-EP-360 EMERGENCY PLAN PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page 1 of 21 4.0 RESPONSIBILITIES 4.1 Upon recognition of EITHER core exit thermocouple temperature(s) > 700 °FOR 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 Engineering Coordinator /TSC Manager of the results of any core damage assessment performed.

5.0 DETAILS NOTE G Core Damage Estimate may be based 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.

o The Core Damage Assessment may be performed as data becomes available. If data is unavailable for a given core damage methodology, then the affected step(s) can be NA'd.

e Contai_nment 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 an increasing 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-2.

5.1 If possible, check or obtain Containment Hydrogen Concentration by either:

a. Dispatching chemistry personnel to obtain sample or
b. Initiate performance of 3-SOP-SS-004, "containment Hydrogen Concentration Measurement System" (Unit 3).

5.2 Determine the possible status of the reactor core using the flowchart in Attachment 1 and perform the associated action.

NON-QUALITY RELATED tabillergy,g.

-.:::::=-

IPEC EMERGENCY PLAN PROCEDURE IP-EP~360 Revision 6 IMPLEMENTING PROCEDURES REFERENCE USE Page .Q of 21 6.0 INTERFACES 6.1 IP-EP-120, Emergency Classification 6.2 EN-EP-610, Technical Support Center 7.0 RECORDS This procedure generates completed Fuel Rod Clad Damage (Attachment 2) and/or Fuel Over-temperature Damage (Attachment 3) worksheets.

8.0 REQUIREMENTS AND COMMITMENTS None 9.0 ATTACHMENTS 9.1 Attachment 1, High Level Core Assessment Flowchart Figure 1A, Containment Radiation Level for 1 %. Fuel Over-temperature 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 Figure 1 B, Containment Radiation Level for 1 %. Fuel Over-temperature Release (>5 hours after shutdown) 9.2 Attachment 2, Fuel Rod Clad Damage 9.3 Attachm~nt 3, Fuel Over-temperature Damage 9.4 Attachment 4, RCS-15

IPEC NON-QUALITY RELATED

~ EMERGENCY PLAN IP-EP-360 Revision 6

~- E'nlergy,. PROCEDURE IMPLEMENTING PROCEDURES REFERENCE USE Page -2 of 21 High Level Core Damage Assessment Flowchart Attachment 1 Start Thermocouples below 700"F NO Containment Radiation YES below 1 R/hr NO Are all Core Exit YES The rm ocou pies below 2000°F Is Containment Possible fuel rod NO Radiation below YES clad dam age, Figure 1A/B go to Attachment 2 Values NO Possible fuel over-

~-------"':=:ao-------, temperature damage, go to Attachment 3

a-=-Entergy~. IPEC EMERGENCY PLAN NON-QUALITY RELATED PROCEDURE IP-EP-360 Revision 6 IMPLEMENTING PROCEDURES REFERENCE USE Page z of 21 Figure 1A Containment Radiation Level for 1% Fuel Overtemperature Release 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 . . - - - - - - - - . - - - - - - - - - - - - - - - - - - - - - ,

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

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

-a--RCS pressure >1600 psig, with containment spray

--+- RCS_pressure <1600 psig, with containment spray 2.50E+03 2.00E+03

.c L ..

-0:::

.ill ro 0:::

Q) f/)

1.50E+03 0

0 co N

---lO N

0:::

1.00E+03 5.00E+02 0.00E+00 .,_______! " ' -_ _ ~----.----------------.-----!

0 1 2 3 4 5 6 7 Time Since Shutdown (hr)

IPEC NON-QUALITY RELATED IP-EP-360 Revision 6 EMERGENCY PLAN PROCEDURE i IMPLEMENTING PROCEDURES REFERENCE USE Page .!! of 21 Figure 18 Containment Radiation Level for 1% Fuel Overtemperature Release

(>5 hours after shutdown)

- RCS pressure >1600 psig, NO containment spray 1.40E+03 ....._ RCS pressure <1600 psig, NO containment spray

- - - RCS pressure >1600 psig, with containment spray

_::+- ~CS p_r_es~~~~ _<_1600 psig, ~j~~.~~~~aJ~.~en~_s_pr~y 1.20E+03

  • 1.00E+03
  • 2 f2. 8.00E+02 (I)

(/)

0 0

co N

i:o 6.00E+02 N

0::

4.00E+02 2.00E+02 -

O.00E+00

. L&Zazwfflffl 1-----~----~-----,-----...------,----__,..

I 0 5 10 15 20 25 30 Time Since Shutdown {hr)

.. ni1lergy~

IPEC NON-QUALITY RELATED PROCEDURE IP-EP-360 Revision 6

,=:::--~ EMERGENCY PLAN IMPLEMENTING PROCEDURES REFERENCE USE Page  !! of 21 Attachment 2 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 2A/BI) A= -----
  • Current containment radiation level (R/hr) B= -----

1.3 Estimate clad damage(%) :

8 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 or SPDS [Unit 3])

  • Number of CETs at or above 1400°F E= - - - - -
  • Number of CETs at or above1200°F F= - - - - -

2.2 For RCS pressure at or above 1600 psigl:

EX 100

% Clad Damage cET = ---- =

D 2.3 For RCS pressure below 1600 psigl:

F x 100

% Clad Damage CET = ----- -

D

IPEC NON-QUALITY RELATED IP-EP-360 Revision 6 EMERGENCY PLAN PROCEDURE i IMPLEMENTING PROCEDURES REFERENCE USE Page 10 of 21 Attachment 2 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)

I I I I 1.80E+04

_ _____________ ! _._ RCS pressure >1600 psig, NO containment spray

___.,_ RCS pressure <1600 psig, NO containment spray

-1 -7*

- - - RCS pressure >1600 psig, with containment spray RCS press'.re _<1600 with_ containment spray 1.40E+04 I

-.... 1.20E+04 I

L:

~

2 m

0::: 1.00E+04

<l.l 11'1 0

0 CD

\

~ 8.00E+03 I{)

N 0:::

\

6.00E+03 4.00E+03 -

2.00E+03 0.00E+00 ~-----,---------~---~----.;.------!

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

a-=-h11lergy~ IPEC EMERGENCY PLAN NON-QUALITY RELATED PROCEDURE IP-EP-360 Revision 6 IMPLEMENTING PROCEDURES REFERENCE Use Page 11 of 21 Attachment 2 Fuel Rod Clad Damage Sheet 3 of 5 Figure 28 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!)

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

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

- - - RCS pressure >1600 psig, with containment spray 3.60E+03 -+- RCS pressure <1600 psig, with containment spray 3.20E+03

.s::

2.80E+03 0::

w* 2.40E+03 ro 0::

Q)

(/)

0 2.00E+03 -

0 (0

N lO N 1.60E+03 0::

1.20E+03 -

8.00E+02 4.00E+02 0.00E+00 0 5 10 15 20 25 30 Time Since Shutdown (hr)

IPEC NON-QUALITY RELATED

,aE11lergy6

-::=-- EMERGENCY PLAN PROCEDURE IP-EP-360 Revision 6 IMPLEMENTING PROCEDURES REFERENCE USE Page 1l of 21 Attachment 2 Fuel Rod Clad Damage Sheet 4 of 5

3. Confirm reasonableness of clad damage estimates.

3.1 Determine the following: (see Attachment 4)

  • 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 0.5f/o? _ _ _ __
  • Is RVLIS between 64f/o and 4f7%?
  • Is hot leg RTD between Tsat and 650°Fp.
  • 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 cErl

% Diff diff = - - - - - - - - - - - - - - - - - - - - - - - - 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
  • Bleed paths from the RCS
  • Direct radiation to the containment radiation monitors
  • r.:,ergy..,

-- II IPEC EMERGENCY PLAN NON-QUALITY RELATED PROCEDURE IP-EP-360 Revision 6 i IMPLEMENTING PROCEDURES REFERENCE USE Page 13 of 21 Attachment 2 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 Manage~.

4.2 Report clad damage estimate to the Engineering Coordinator/TSC Manage~.

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

=* E.'nlergy..,

IPEC EMERGENCY PLAN IMPLEMENTING PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE IP-EP-360 Page 14 Revision 6 of 21 Attachment 3 Fuel Over-temperature 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 3AfiB) 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).

2.1 Determine the following :

(Refer to PIGS [Unit 3])

  • Number of CETs at or above 2000°IF K= - - - - -

2.2 Estimate fuel overtemperature damage(%):

Kx 100

% Core Damage cET = - - - =

J

e.

=-ti1lergy,,

IPEC EMERGENCY PLAN NON-QUALITY RELATED PROCEDURE IP-EP-360 Revision 6 i IMPLEMENTING PROCEDURES REFERENCE USE Page 15 of 21 Attachment 3 Fuel Over-temperature Damage Sheet 2 of 7 Figure 3',A 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) 3.00E+05 - - - - - - - - _ -- - - - - - - - - - - - - - - - - - - - - - - - - .

-e- RCS pressure >1600 psig, NO containment spray

__._ RCS pressure <1600 psig, NO containment spray

- - - RCS pressure >1600 psig, with containment spray

--+- RCS pressure <1600 psig, with containment spray __

2.50E+05

.c 2.00E+05

~

-Q) co a::

Q) 1/)

1.50E+05 0

0 CD

--a::

N LO N

1.00E+0S -

5.00E+04 -

0.00E+00 ~--=-------.,..----.. . . ----.,..----.. . . ------f 0 1 2 3 4 5 6 Time Since Shutdown (hr)

=-t:11tergy~

IPEC EMERGENCY PLAN IMPLEMENTING PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE IP-EP-360 Page 16 Revision 6 of 21 Attachment 3 Fuel Over-temperature Damage Sheet 3 of 7 Figure 3 S* II Containment Radiation Level for 100% Fuel Overtemperature Release

(>5 hours after shutdown) l

-e- RCS pressure >1600 psig, NO containment spray 1.40E+05 ------- -- --------- ----! _..,_ RCS pressure <1600 psig, NO containment spray

-a- RCS pressure >1600 psig, with containment spray L -+- RCS pressr ~"'"~ psi~- ~th containment spray 1.20E+05 1.00E+05

'L:'

.c

~

Q) ro

~ 8.00E+04 Q) rJ'l 0

0 c.o N

in N

~

6.00E+04 4.00E+04 2.00E+04 0.00E+0O .

0 5 10 15 20 25 30 Time Since Shutdown (hr)

-===- Entergy.!,

IPEC EMERGENCY PLAN IMPLEMENTING PROCEDURES NON-QUALITY RELATED PROCEDURE REFEREN,CE USE IP-EP-360 Page 17 Revision 6 of 21 Attachment 3 Fuel Over-temperature 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 105Q Yes CH2j No CH~

Atorabove105Q Yes CH4j No CH~

3.2 Estimate fuel overtemperature damage (%):

L X 100

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

M

@E11Jer.oy. b 1' IPEC EMERGENCY PLAN IMPLEMENTING NON-QUALITY RELATED PROCEDURE IP-EP-360 Revision 6 PROCEDURES REFERENCE USE Page 18 of 21 Attachment 3 Fuel Over-temperature 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 - - - - - - - - - - - - - - - - - - - - - L-:__

___-__... - ...1- - . . . - -...

Ij - - -C H2

CH2 dry wet I!

9 +--- - - - - -

- - -- - -- -- ---, - -

  • CH3wet

- - --- - --t ::::~~::I

--+--CH3 dry 8 --------

R 7 -*

~

C:

~

~

c 6

~

C:

0 u

C:

Q)

Cl 5-e

"'O .

c c

'I **

Q) 4 llt-o;;;;;;;;;:fl"---t:t---li't---"~--"-ii!f-----!D--"-----------'-il--lr-----=fl E

C:

. l'. *

~

C:

0 u

i

.. " I i ..

2 .. t

.... +.

- - ,. . I

- - - L ....

I 1 I 0 +---,----,.---;----------~~-+---+---+-~

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

-=

-=- hi1lergy~

IPEC EMERGENCY PLAN NON-QUALITY RELATED PROCEDURE IP-EP-360 Revision 6 i IMPLEMENTING PROCEDURES REFERENCE USE Page 19 of 21 Attachment 3 Fuel Over-temperature 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 hot leg RTD at or above 650°Fj?
  • 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

% Diff diff - ___________________________________________,________ x 100

  • % 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 diff - --------------------------------------------------- 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 o/o Diff diff = ---------------------------- X 100

% Core Damage Hyd

e*=- E'nlergy., IPEC EMERGENCY PLAN NON-QUALITY RELATED PROCEDURE IP-EP-360 Revision 6 i IMPLEMENTING PROCEDURES REFERENCE Use Page 20 of 21 Attachment 3 Fuel Over-temperature 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 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 Engineering Coordinator/TSC Manag~r.
6. Return to Step 5.1 of this procedure to continue assessment of the reactor core.

.e

=H11tergy.,

IPEC EMERGENCY PLAN NON-QUALITY RELATED PROCEDURE IP-EP-360 Revision 6 IMPLEMENTING PROCEDURES REFERENCE USE Page 21 of 21 Attachment 3 RCS-15 Sheet 1 of 1 RCS-15, Rev. 0 RVLIS Full Range Level Indication Map 100 75 90 70 80 65' 70

~

........... 60 60

z .........

0 ----------- .. *----------- s

~

c...>

z 0

50 a 55

~

~

en

~

i et::

40 CORE 50

.....J w

30

. ------------ 45 20 40 10 0 .35 RVLIS Approximate Vessel Elevation Indication,. Water Level 100% Top of Vessel 76.5 ft 82% Vessel Flange 69 ft 667o Inlet/Outlet Nozzles 62 ft 56% Top of Core {UCP) 57.8 ft 24% Bottom of Core (LCP) 44.5 ft 0% Bottom of Vessel 34.1 ft

  • These values do not include harsh environment. uncertainty of 6%

Written by*, .,::.=~....!.:::~::+:=:.......--

Reviewed by:~~~~~~~6..-

PORC Review: 9: '.Z;.

  • Approved by: _ s/;/ ~

Effective Date: 3l-r9{,/(i- 9:3__

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