ML082960531

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Revision 4 to Emergency Preparedness Procedure 10-S-01-35, Core Damage Assessment.
ML082960531
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 10/16/2008
From:
Entergy Operations
To:
Document Control Desk, Office of Nuclear Security and Incident Response
References
10-S-01-35, Rev 4
Download: ML082960531 (19)
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Text

GRAND GULF NUCLEAR STATION EIE SUBMISSION DATE October 22, 2008 TO DOCUMENT CONTROL DESK FROM GRAND GULF NUCLEAR STATION LICENSE NUMBER NPF-29 DOCKET NUMBER 50-416 TITLE GRAND GULF NUCLEAR STATION EMERGENCY PREPAREDNESS PROCEDURE 10-S-01-35, REVISION 4, CORE DAMAGE ASSESSMENT NUMBER OF PAGES 19 FORMAT PDF - SEARCHABLE IMAGE (EXACT)

RESOLUTION 300 DPI REGULATION GOVERNING 10CFR50.4(b)(5)

SUBMISSION SUBMISSION CONTACT ANN M MARSHALL PH 601-437-6486 INFORMATION EMAIL: ATOWNSE@ENTERGY.COM

PLANT OPERATIONS MANUAL Volume 10 10-S-01-35 Section 01 Revision: 004 Date: /()-/&r-tJ<%

REFERENCE USE EMERGENCY PLAN PROCEDURE CORE DAMAGE ASSESSMENT SAFETY RELATED prepared:

Reviewed:

Approved:

List of Effective Pages:

Pages 1-7 Attachment I-VI List of TCNs Incorporated:

Revision o -004 None J:\ADM_SRVS\TECH_PUB\REVISION\10S0135.DOC

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE Tltle: Core Damage Assessment No.: 10-S-01-35 Revlslon: 004 Page: i REQUIRED REVIEW PERFORMED ~PAD --- D 50.59 Evaluation (Check all that apply) (EN-LI-100) (EN-LI-101)

D 72.48 Evaluation ~ 50.54 Evaluatlon Transmit applicable Review Form as (EN-..L I-112 ) (ENS-NS-210) a separate record along with ~PAD Not Required(EN-LI-100 or 01-S-02-3) procedure to Document Control. D Process Applicability Excluded D Editorial Change D lSI/1ST Implementation

~N Incorporatlon or Auto R~. ~

Other Process-Number: cc.. 70 PAD Revlewer  :~tOJ!~ //{)J5/P~

(for PAD Not Required) Slqnature/Date Cross-Discipline review required? (Note affected Departments Below)

( )Yes (X) No_

I Preparer Initials>>> J

'/

Department Cross-Discipline Reviews Needed Signoff (signed, electronic, telcon)

Does this directive contain Tech Spec Triggers? ( ) YES (X) NO J:\Adm_srvs\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Core Damage Assessment No.: 10-S-01-35 Revision: 004 Page: il REQUIREMENTS CROSS-REFERENCE LIST Requirement Implemented by Directive Directlve Paragraph Number Name Paragraph Number That Implements Requirement NEDC-33045P+ 5.0 *

+ - Endorsed by the NRC's Safety Evaluatlon dated June 12,2001, for BWROG Toplcal Report NEDO-32991 as referenced in GGNS License Amendment 158.

  • - Covered by directive as a whole or by various paragraphs of the directive.

NOTE The Equipment Database (EDB) Request statement is applicable only to volume 06 and 07 maintenance dlrectives.

EDB Change Request generated and the h~kUP documentation available for setpoint and/or calibratlon data only D Yes ~;A EDBCR #

Current Revision Statement Revision 004:

  • Changed the MZIRWL in Section 6.2.1 and Attachment I from -212 inches to

-210 inches in accordance with Calculation XC-Q1111-99001, EP/SAP Calculations, Revision 7.

J:\Adm_srvs\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Core Damage Assessment No.: 10-S-01-35 Revision: 004 Page: 1 1.0 PURPOSE The purpose of this procedure is to provide a simple qualitative evaluation of the overall status of the reactor core under severe acc1dent cond1tions. The results of the evaluation support appropriate mitigative and protect1ve actions while an event is 1n progress.

2.0 RESPONSIBILITIES 2.1 Accident Assessment Engineer - Is responsible for the performance of this procedure, core damage estimate calculations, and providing core damage assessment to the Offsite Emergency Coordinator.

2.2 TSC Coordinator - Is responsible for monitoring plant conditions for ind1cat1on of core damage.

3.0 REFERENCES

3.1 NEDC-33045P, "Methods of Estimating Core Damage in BWRS" , Revis10n 0, General Electric Co., July 2001.

3.2 RTM-96, "Response Technical Manual", Volume 1, Revision 4, USNRC, March 1996.

3.3 Calculation XC-Qllll-99001, Plant Specific Technical Guidelines, Calculation for Plant-Specific Var1ables and Graphs, Rev1sion 5.

3.4 05-S-01-EP-2, "RPV Control".

3.5 Calculation JC-QIE61-R602-1, DrYWell and Conta1nment Hydrogen Concentration Loop Uncertainty, Rev1sion O.

4.0 ATTACHMENTS 4.1 ATTACHMENT I - Core Cooling Assessment 4.2 ATTACHMENT II - DrYWell Radiation Levels 4.3 ATTACHMENT III - Containment Radiation Levels 4.4 ATTACHMENT IV - Volume-Weighted Average Hydrogen Concentration 4.5 ATTACHMENT V - Zirconium Oxidation Fraction 4.6 ATTACHMENT VI - Core Damage Tracking Data 5.0 DEFINITIONS 5.1 CORE DAMAGE - Widespread degradation of the fuel pellet or cladding fission product barriers due to inadequate core cooling. Core damage is characterized by the combinat1on of cladding and overheating damage.

5.2 NO DAMAGE - A condition in which there is no positive indication of core damage has been detected. Fuel cladding temperatures below 1500 0 F and fission product releases are limited to the radionuclides normally present in the reactor coolant 5.3 CLADDING DAMAGE - Degradation of the fuel rod cladding permitting the release of gaseous and volatile fission products from the space between the fuel pellets and the fuel cladding. Sign1ficant cladding damage 1S expected to begin at temperatures of approximately 1500 0 F.

J:\Adm_srvs\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Core Damage Assessment No.: 10-S-01-35 Revision: 004 Page: 2 5.4 OVERHEATING DAMAGE - Damage to the fuel from temperatures high enough to release gaseous and volatile fission products from the fuel matrix but insufflclent to melt the fuel.

5.5 FUEL MELT - Damage that occurs at higher temperatures and results in changes to the core geometry. Thls phase of core damage is not addressed by this procedure since no mitigative actions are based on the onset of melting and it is not expected that the condition would be easily detectable by plant personnel.

5.6 Minimum Steam Cooling RPV Water Level (MSCRWL) - The lowest RPV water level at which the covered portion of the reactor core will generate sufflcient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500 0 F (-191 inches).

5.7 SPIKED COOLANT RELEASE - The release into containment of 100 times the non-noble gas fission products normally found in the coolant. Spikes are normally associated with a rapid shutdown or depressurization.

6.0 DETAILS 6.1 General Directions 6.1.1 The type and extent of core damage are estimated based on assessments of core cooling history, primary containment radlation, and primary containment hydrogen concentration. The initial screening identifies conditions in which core damage may be possible. Primary containment radiation levels provide the primary indicator wlth core cooling and hydrogen concentration providing confirmatory indications.

NOTE Interpolate or extrapolate from the predicted levels given in Attachments II & III to account for decay between sub-criticality and when readinqs are taken.

6.1.2 Check for indication of possible core damage using Table 1 below.

Make adjustments to the predicted radiation levels, as necessary.

Table 1 Indication of possible Core Damage Condition Indications Core Coollng Assessment See Attachment I Primary containment radlatlon. DrYWell and Contalnment radiatlon levels above the maximum predicted coolant release values for the Spiked Coolant release - See Attachments II and III.

Hydrogen generation DrYWell or containment hydrogen concentration above 0.8%.

6.1.3 The core damage assessment is expected to be performed many times during the course of an accident. Use Attachment VI to record the results of these assessments.

6.1.4 If core damage is indicated, evaluate the type and amount of damage using Table 2 below.

J:\Adm_srvs\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Core Damage Assessment No.: 10-S-01-35 Revislon: 004 Page: 3 NOTE Evaluate drywell and contalnment radiation levels independently.

Table 2 Preliminary Core Damage Assessment Drywell/Containment Possible Damage States Action Radiation

{Attachments I I and III}

Below 1% Gap Release Core damage unlikely None (cladding failure}

radiation levels Between 1% Gap Release Limited claddlng Estimate the amount of (cladding failure} damage cladding damage using radiation levels and 1% and Section 6.2.

Overheating Release Overheating damage radiation levels. unlikely Above 1% Overheating Widespread cladding Estimate the amount of Release radiation levels. damage cladding damage using and Section 6.2.

possible overheating and damage Estimate the amount of overheating damage using Section 6.3.

Above 100% Gap Release Widespread cladding Verify the amount of (cladding failure) damage cladding damage using radiation levels. and Section 6.2.

Widespread overheating and damage Estimate the amount of overheating damage using Section 6.3.

6.2 Estimating Cladding Damage 6.2.1 Indications

a. DrYWell or containment radiation level above the 1% Gap Release (cladding failure) radiation level.
b. DrYWell or containment hydrogen concentration above 0.8%.
c. RPV water level below MSCRWL {-191 inches}
d. RPV water level below MZIRWL (-210 inches) without RPV injection
e. Inadequate core spray (HPCS and LPCS flow less than design

{7000 gpm and RPV water level below -217 inches. 6.2.2 Procedure

a. Determine the predicted drYWell and containment radiation levels corresponding to 100% Gap Release {cladding failure}

from Attachments II and III using an appropriate tlme adjustment. J:\Adrn_srvs\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Core Damage Assessment No.: 10-5-01-35 Revlsion: 004 Page: 4 6.2.2 (Cont. )

b. Estimate the amount of cladding damage as follows:

NOTE Perform separate calculatlons for the drywell and containment. Use the higher of the two results. Indicated Radiation Level 

             % Cladding Damage     -------------xlOO 100 % Gap Re lease Radiation Level NOTE If igniter operatlon or other hydrogen depleting mechanisms have occurred (see 7.3.3), Attachment IV and V will underestimate the amount of zirconium oxidation.
c. If Hydrogen is present, estimate the amount of zirconium oxidation using Attachment IV or V.
d. Refer to Section 7 for guidance on interpreting the damage assessment results.

0-10% damage = Limited Damage 

                             >10% damage  = Widespread  Damage 6.3    Estimating Overheating Damage 6.3.1     Indications
a. Drywell or containment radiation level above the 1%

Overheating Release radiation level. (Attachments II and III) .

b. Drywell or containment hydrogen concentration above 2.0%.
c. RPV water level below -217 inches and Core Spray flow less than design (7000 gpm) for more than 30 min.

6.3.2 Procedure

a. Determine the drywell and containment radiation levels corresponding to 100% Overheating Release from Attachments II and III.
b. Estimate the amount of overheating damage as follows:

NOTE Perform separate calculations for the drywell and containment. Use the higher of the two results. J:\Adm_srvS\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE Tltle: Core Damage Assessment No.: 10-S-01-35 Revision: 004 Page: 5 6.3.2 (Cont. ) m 0 h . D IndicatedRadiationLevel 100 

                 -/0   ver eatzng amage =                                         X 100% Overheating Re leaseRadiationLevel NOTE If igniter operation or other hydrogen depleting mechanisms have occurred (see 7.3.3). Attachment IV and V will underestimate the amount of zirconium oxidation.
c. Estimate the amount of zirconium oxidation using Attachment IV or V. Use Attachment V if a more accurate estimate lS needed.
d. Refer to Section 7 for guidance on interpreting the damage assessment results.

0-10% damage Limited Damage 

                                >10% damage   Widespread Damage 7.0  DISCUSSION 7.1    General 7.1.1      This procedure provides a general evaluation of the status of the core. The accuracy of the results is governed by source term uncertainties, instrumentation characteristics, and event-specific variations.

7.1.2 Use appropriate engineering judgment to reconcile any differences between assessment results and obtain a best estimate of core damage. 7.1.3 Due to the power distribution in the core during operation, core damage is expected to occur at different rates in different parts of the core. Overheating damage may exist in the hottest regions of the core while fuel in lower power regions remains undamaged. 7.2 Radiation Levels (Attachments II and III) 7.2.1 A radiation level equivalent to 1% of the gap release inventory is positive indication of cladding damage. 7.2.2 The cladding damage and overheating damage radiation ranges overlap. Withln the overlap range, the relative amounts of each type of damage cannot be distinguished. Calculating separate values for each type of damage tends to overestimate both amounts. 7.2.3 A radiation level above the 1% overheating damage value is not positive indlcation of overheating damage since the same radiation level could be reached with only cladding damage. However, it is unlikely that widespread cladding damage would exist without at least some overheating damage. Above the 50% cladding damage radiation level, it may be assumed that overheating damage also exists. J:\Adm_srvs\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Core Damage Assessment No.: 10-S-01-35 Revision: 004 Page: 6 7.2.4 The relative values of drywell and containment radiation levels are event-dependent. If a primary system break exists and steam is discharged into the drywell, drywell radiatlon levels will probably increase first. If no primary system break exists and steam is discharged through the SRVs, containment radiation level will probably increase flrst. 7.3 Hydrogen 7.3.1 Rapid hydrogen production begins at temperatures above the threshold for cladding damage but below the onset of significant overheating damage. 7.3.2 Detectable hydrogen is an indication of possible cladding damage. 7.3.3 Hydrogen removal mechanisms including igniter operation in the drywell or containment, spurious hydrogen burns, or containment venting are expected to significantly reduce the hydrogen concentration and reduce the containment oxygen concentration. Since containment oxygen concentrations are not available, it is impossible to assess the impact of these removal mechanisms on the Hydrogen concentration. As a result, hydrogen concentration can only be correlated to core damage when igniter operation or other removal mechanisms have not occurred. 7.3.4 If no hydrogen is detected, it is unlikely that overheating damage exists. 7.4 Uncertainties 7.4.1 Radiation level measurements may underestimate core damage if:

a. The primary containment or RPV has been vented.
b. primary system isolations have been defeated to permit continued use of the main condenser under failure-to-scram conditions.
c. Primary containment integrity has been lost.

7.4.2 Radiation level measurements may overestimate core damage if:

a. The suppression pool has been bypassed.
b. Suppression pool water level is low.

7.4.3 Hydrogen concentration measurements may underestimate core damage if:

a. Hydrogen igniters have operated or a spurious hydrogen burn has occurred.
b. The primary containment has been vented.
c. Primary containment integrity has been lost.
d. Significant amounts of hydrogen remain trapped in the RPV.

J:\Adm_srvS\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE

Title:

Core Damage Assessment No.: 10-S-01-35 Revision: 004 Page: 7 7.4.4 Hydrogen concentration measurements may over estimate core damage if:

a. Significant amounts of hydrogen have been generated by radiolysis.
b. The hydrogen injection system is leaking.
c. Steam is present in the drywell but the drywell atmosphere is not at saturation conditions.

J:\Adm_srvs\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-35 Revl.sl.on: 004 Attachment I Page 1 of 1 START 

                            >------No>----~~~IL.._ _N_o_c_o_nc_lu_s_io_n_ _

Yes No Yes Yes Yes No 

                      ,_-------Ye's--------------<

No Attachment 1: Core Cooling Assessment J:\Adrn_srvS\TECH_PUB\REVISION\lO\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-35 Revision: 004 Attachment II Attachment II Page 1 of 1 Drywell Radiation Levels Spiked Gap Release Gap Release Normal Normal Coolant Spiked (Cladding (Cladding Overheating Overheating Coolant Coolant Release Coolant Failure) Failure) Release Release Release 1hr Release 24 hr 1 hr Release 24 hr 1hr 24hr 1 hr 24 hr 1.0E 1.0E-03 1.0E-04 !~~~~ 1.0E-05 -'-- -'- -'- --'- ----' ..1.-- -'- -'- --' Shaded areas represent 1% (low value) to 100% (high value) ofinventory release to the drywell J:\Adrn_srvs\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-35 Revision: 004 Attachment III Attachment III Page 1 of 1 Containment Radiation Levels Spiked Gap Release Gap Release Normal Normal Coolant Spiked (Cladding (Cladding Overheating Overheating Coolant Coolant Release Coolant Failure) Failure) Release Release Release 1hr Release 24 hr 1 hr Release 24 hr 1hr 24hr 1 hr 24 hr a; 

> 1.0E+02 CIl

...I c:: 0

1.0E+01

't' C'Cl II: 1.0E+OO ...c CIl E 1.0E-01 c ...c C'Cl 0 1.0E-02 0 1.0E-03 Shaded areas represent 1% (low value) to 100% (high value) of inventory release to the containment J:\Adm_srvs\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-35 ReVlSlon: 004 Attachment IV Page 1 of 2 Attachment IV: Volume-Weighted Average Hydrogen Concentration Use this worksheet to calculate the volume-weighted average containment hydrogen concentration. Drywell hydrogen concentration (%): ~w =  % Containment hydrogen concentration (%): Hsc = - - - - - - - - _ % Drywell airspace free volume (fe): V dw = 270,000 fe Containment airspace free volume (fe): 1,400,000 fe Volume-weighted average hydrogen concentration: H = H dw (270,OOO) + H sc (1.4E06) AVG 1.67 E6 =--------- % Determine % Zr Oxidation from the graph on page 2 of this attachment. J:\Adm_srvs\TECH_PUB\REVISION\10\lOS0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-35 ReVlSlon: 004 Attachment IV Page 2 of 2 Zirconium Oxidation Fraction 100 90 / 80 v 

                                                                                /

70 / 

             "C (1)
             "~
             "C       60
                                                                     /
                                                               /
             "S<

a E 50 V

I C

0 0

to-N 40
             'if.

30 

                                              /
                                                 /

20 / 10 /' V 0 / o 5 10 15 20 25 30 Average Hydrogen Concentration (From Attachment IV, Page 1) Attachment IV: Volume-Weighted Average Hydrogen Concentration J:\Adm_srvs\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-35 ReVlSlon: 004 Attachment V Page 1 of 1 Attachment V Zirconium Oxidation Fraction Use this worksheet to verify estimates obtainedfrom Attachment IV.

1. Calculate the mass of oxidized zirconium based on the drywell hydrogen concentration:

Drywell hydrogen concentration (%): H dw = - - - - - - - - _ % Drywell airspace free volume** (ft3): 270,000 fe Drywell pressure (psig): Pdw = psig Drywell temperature (OF): Tdw= OF dw Saturation pressure for Tdw (psia): Psal _ _ _ _ _ _ _ _ _psia Mass of oxidized zirconium in the drywell: 

                                                                                     =                             lbm
2. Calculate the mass of oxidized zirconium based on the containment hydrogen concentration:

Containment hydrogen concentration: He=  % Containment airspace free volume**: 1,400,000 ft 3 Containment pressure Pc = psig Containment temperature Tc = oF Saturation pressure for T P,~I = psia Mass of oxidized zirconium in the containment: mC = 0.04242 X H C(1.4E06)(Pc + 14.7 - P,~I) = lbm Zr (460+TJ

3. Calculate the total zirconium oxidation fraction:

dw c F = mZr + mZr xl00 --------_% Zr 71007

    • - Values based on normal water lllventory. Adjust to account for significant flooding of either the drywell or containment.

J:\Adm_srvS\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-35 Revision: 004 Attachment VI Page 1 of 2 Attachment VI Core Damage Tracking Data DATE:, _ Cladding Damage Time Cladding 100% Gap Release Radiation levels (R/hr) Cladding Damage Hydrogen Zirconium Overall Damage (24 hr) Damage (cladding failure) DW/Cont Assessment (%) Concentration (%) Oxidation (%) Estimate (%)1 Indicated? Radiation Level (6.2.2.b) DW/Cont (6.2.2.c) (6.2.2.d) (6.2.1) (6.2.2.a) DW/Cont YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / 1 0-10% damage =Limited Damage; >10% damage =Widespread Damage J:\Adrn_srvs\TECH_PUB\REVISION\10\10S0135.doc

GRAND GULF NUCLEAR STATION EMERGENCY PLAN PROCEDURE 10-S-01-35 Revision: 004 Attachment VI Page 2 of 2 Attachment VI Core Damage Tracking Data DATE:. _ Overheating Damage Time Overheating 100% Damage Radiation levels (Rlhr) Overheating Hydrogen Zirconium Oxidation Overall Damage (24 hr) Damage Radiation Level DW/Cont Damage Concentration (%) (%) Estimate (%)1 Indicated? (6.3.2.a) Assessment (%) DW/Cont (6.3.2.c) (6.3.2.d) (6.3.1) DW/Cont (6.3.2.b) YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / YES / NO / / / 1 0-10% damage = Limited Damage; > 10% damage = Widespread Damage J:\Adm_srvs\TECH_PUB\REVISION\10\10S0135.doc}}

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