ML20195E767

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Rev 0 to EPP 01-2.4, Core Damage Assessment Guidance
ML20195E767
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 11/10/1998
From:
WOLF CREEK NUCLEAR OPERATING CORP.
To:
Shared Package
ML20195E756 List:
References
EPP-01-2.4, EPP-1-2.4, NUDOCS 9811190107
Download: ML20195E767 (15)
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l ENCLOSURE 3 l

EPP 01-2.4, CORE DAMAGE ASSESSMENT GUIDANCE, DRAFT REVISION O 9811190107 981110 i PDR ADOCK 05000402 P PDRj l

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W6LF CREEK ' NUCLEAR OPERATING CORPORATION EPP 01-2.4 CORE DAMAGE ASSESSMENT GUIDANCE Responsible Manager Manager Nuclear Engineering Revision Number DRAFT Use Category Reference Administrative Controls Procedure No Infrequently Performed Procedure Yes Program Number 06

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Revision: DRAFT CORE DAMAGE ASSESSMENT GUIDANCE EPP 01-2,4 Reference Use Page 1 of 13 TABLE OF CONTENTS SECTION TITLE PAGE i

1.0 PURPOSE 2 2.0 SCOPE 2

3.0 REFERENCES

AND COMMITMENTS 2 l 4.0 DEFINITIONS 3 5.0 RESPONSIBILITIES 3 6.0 PRECAUTIONS / LIMITATIONS 3 )

7.0 PROCEDURE 5 8.0 INITIAL ACTIONS 10 9.0 SUBSEQUENT ACTIONS 10 10.0 RECORDS 10 11.0 FORMS 10 FIGURE 1 CONTAINMENT DOSE RATE RESPONSE TO TECHNICAL SPECIFICATION 3.4.8 SOURCE TERM WITH PRE-EXISTING IODINE SPIKE 11 FIGURE 2 CONTAINMENT DOSE RATE RESPONSE TO GAP RELEASE SOURCE TERM 12 FIGURE 3 CONTAINMENT DOSE RATE RESPONSE TO CORE OVERTEMPERATURE SOURCE TERM 13 s

Re, vision: DRAFT CORE DAMAGE ASSESSMENT GUIDANCE EPP 01-2.4 6

Reference Use Page 2 of 13 1.0 PURPOSE

- 1.1. This guideline provides information for the assessment of the degree of core damage during an accident. (Commitment Step 3.2.1]

1.2 In addition, the guideline provides information for the ,

assessment of the appropriate Emergency Action Level for off-  !

site radiological protective actions based on the degree of core damage. Specifically, the information contained in this guideline relates to:

  • Determination 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, ,
  • . Determination of the degree of core overheating that results in the release of the fission product inventory in the fuel i pellets, and
  • Determination of the appropriate Emergency Action Level for off-site radiological protective actions, based on the degree of damage to the reactor core. l

'2.0 SCOPE 2.1 This procedure is applicable to the Engineering Team in the

. Technical Support Center (TSC) for use during a declared emergency.

, 2.2 This procedure is applicable whenever there are indicated core temperatures that trigger the use of the Functional Restoration )

Guidelines or whenever there are indicated containment radiation levels that trigger an alarm.

3.O REFERENCES AND CO60GTMENTS 3.1 References 3.1.1 WCAP-14696, Westinghouse Owner's Group Core Damage Assessment Guidance, July 1996 i 3.1.2 NE 98-0090, Setpoints for WCAP-14696 Core Damage Assessment Guidance 3.1.3 Nuclear Engineering Calculation AN 98-029 Revision 0 Containment High-Range Area Monitor Radiation Level Determination for Updating Core Damage Assessment Guidance

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Revision: DRAFT CORE DAMAGE ASSESSMENT GUIDANCE EPP 01-2.4 Reference Use Page 3 of 13 3.2 Commitments 3.2.1 Commitment 85-401, IR 8540-04, EPPs Will Include Core Damage Assessment Procedure 4.0 DEFINITIONS 4.1 No Core Damage 4.1.1 A core state in which the integrity of the fuel rod cladding is intact and the only release of fission products to the reactor coolant system is that due to pre-existing fuel rod defects and iodine spiking.

(Reference Step 3.1.1) 4.2 Fuel Rod Clad Damage 4.2.1 A core state in which the fuel rod cladding of some fraction of the fuel rods in the core has failed, resulting in the release of the fission products in '

the fuel rod gap space of the failed fuel rods to the reactor coolant system. (Reference Step 3.1.1) 4.3 Fuel Over-temperature Damage 4.3.1 A core state in which the fuel pellets have reached a temperature where there is a rapid movement of fission products from the fuel pellet matrix to the reactor ,

coolant system. (Reference Step 3.1.1) J 4.4 100% Fuel Rod Clad Damage 4.4.1 The rupture of the fuel rod cladding in 100% of the ,

fuel rods in the core and the resultant release to the  !

reactor coolant system of all fission products l contained in the fuel rod gap space. (Reference Step  !

3.1.1) l 5.0 RESPONSIBILITIES 5.1 TSC Engineering Team Nuclear Engineer 5.1.1 Assess the degree of fuel damage in accordance with this procedure.

6.0 PRECAUTIONS / LIMITATIONS 6.1 The distribution of airborne isotopes is assumed to be homogeneous throughout the containment free volume.

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9 Revision: DRAFT CORE DAMAGE ASSESSMENT GUIDANCE EPP 01-2.4 Reference Use Page 4 of 13 6.2 Figures 1, 2, and 3 are intended to be representative or best estimate, rather than conservative or bounding values. The results can be affected by:

  • Uncertainties in total core inventory,
  • Fractional migration of firsion products from the core to the containment,
  • Integrity of the fuel matrix above 40 GWD/MTU,
  • Changes in the density of the containment atmosphere,
  • Fission product plateout during accident conditions, and
  • Fission product accumulation in operating equipment.

6.3 Detectors GTRE0059 and 0060 view different size and locations of air volumes within containment. It is anticipated that physical processes during severe accident conditions will cause higher levels of radioactivity in certain regions of containment. For example, Atmosphere Control Filter Adsorbers FGR01A/B are located in close proximity to GTRE0059. Rainout of fission products may collect on the floor areas surrounding GTRE0060 on 2047' elevation. Both of these situations would increase the detector's indication.

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Revision: DRAFT CORE DAMAGE ASSESSMENT GUIDANCE EPP 01-2.4 Reference Use Page 5 of 13 7.0 PROCEDURE 7.1 Identify Current Plant Status.

7.1.1 Using the table below, determine the possible status of the reactor core.

7.1.2 Go to the appropriate section of this procedure as l indicated from the table.

l High Level Core Damage Assessment Fuel Rod Plant Status Fission Product 1

Status )

1 1 1 Core Exit Thermocouples LESS THAN 712 *F No Core Damage l AND l l

Containment Radiation LESS THAN Figure 1 R/Hr Continue to Monitor Plant Parameters Core Exit Thermocouples LESS THAN 2000 *F Possible Fuel Rod AND EITHER Clad Damage j With RCS Pressure GREATER THAN 1600 psig; Containment Radiation LESS THAN 1% of Figure 3 R/Hr Go to Step 7.2. l (0.01

  • R/Hr]

SLR i

With RCS Pressure LESS THAN 1600 psig; I

! Containment Radiation LESS THAN 1% of Figure 3 R/Hr

[0.01

  • R/Mr]

Core Exit Thermocouples GREATER THAN 2000 *F Possible Fuel OR EITHER Over-temperature With RCS Pressure GREATER THAN 1600 psig; Damage l

Containment Radiation GREATER THAN 1% of Figure 3 R/Hr (0.01

  • R/Hr] Go to Step 7.3.

l O.R l With RCS Pressure LESS THAN 1600 psig; l

Containment Radiation GREATER THAN 1% of Figure 3 R/Hr (0.01

  • R/Hr]

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Revision: DRAFT CORE DAMAGE ASSESSMENT GUIDANCE EPP 01-2.4 Reference Use Page 6 of 13 7.2 Fuel Rod Clad Damage 7.2.1 Estimate Fuel Rod Clad Damage Based on Containment Radiation Levels.

1. Find containment radiation level for 100% clad damage.from Figure 2.
2. Obtain current containment radiation level.
3. Estimate clad damage using:

Current Containment Radiation Level

% Clad Damage *' ' "

Predicted Containment Radiation Level at 100% Clad Damage  :

, 1 4

7.2.2 Estimate Fuel Rod Clad Damage ~ Based on Core Exit 1 Thermocouple Readings.  !

1. With RCS Pressure GREATER THAN 1600 psig: {

1 Number ofCETs > 1600* F

% Clad Damage" "

TotalNumber ofOperable CETs

2. ' With RCS Pressure LESS THAN 1600 psig:

Number ofCETs > 1200*F

% Clad Damage" "

TotalNumber ofOperable CETs l

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Revision: DRAFT CORE DAMAGE ASSESSMENT GUIDANCE EPP 01-2.4 Reference Use Page 7 of 13 7.2.3 Confirm Reasonableness of Clad Damage Estimates.

1. Compare to expected response.
  • Containment Hydrogen Concentration LESS THAN 0.2 (DRY) volume percent
  • RVLIS LESS THAN 2019.7' AND GREATER THAN 2004.4'
  • Hot Leg RTD GREATER T t AND THAN LESS THAN 650 F
  • Source Range Monitor GREATER THAN 8000 Counts per Second
  • Difference in clad damage astimates from containment radiation (CRM) and core exit thermocouples (CET) LESS THAN 50%, using

. % Clad Damage *. - % Clad Damage *. _

ABSOLUTE VALUE

% Clad Damage, _

2. If expected response is not obtained, determine if the deviation can be explained from the accident progression.
  • Injection of water to the RCS
  • Bleed paths from the RCS
  • Direct radiation to the containment radiation monitors 7.2.4 Report clad damage estimate to the Technical Support Center Engineering Coordinator.
1. If clad damage estimates have increased by more than 1% in the past 30 minutes or if estimates exceed 5% clad damage, report possible change in Emergency Action Level to the Technical Support Center Engineering Coordinator.

7.2.5 Return to Step 7.1 in Diagnostic Section.

Ravision: DRAFT CORE DAMAGE ASSESSMENT GUIDANCE EPP 01-2.4 Reference Use Page 8 of 13 7.3 .yuel Over-temperature Damage' 7.3.1- Estimate Fuel Over-temperature Damage Based on Containment Radiation Levels.

1. Find containment radiat a level for 100% over-temperature damage from Figure 3. I
2. Obtain current containment radiation level.

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3. Estimate over-temperature damage using:

Current Ctmt Radiation Level

% Core Damage". ' "

Predicted Ctml Radiation Level at 100% Overtemp Damage 7.3.2 Estimate Fuel Over-temperature Damage Based on Core Exit Thermocouple Readings.

-1. Obtain current core exit thermocouple temperature readings.

2. Estimate over-temperature damage using:

Number ofCETs > 1600* F

% Core Damage" "

TotalNumber ofOperable CETs 7.3.3 Confirm Reasonableness of Fuel Over-temperature Damage Estimates.

1. Compare to expected response
  • RVLIS LESS THAN 2004.4' 3
  • Hot Leg RTD GREATER THAN 650 F
  • Source Range Mc,nitor GREATER THAN 8000 Counts ,

Per Second

  • Difference in fuel over-temperature estimates from containment radiation and core exit thermocouples LESS THAN 50%, using:

% Core Damage"' - % Core Damage *.

ABSOLUTE VALUE

% Core Damage m, _

e Containment hydrogen cont .tration

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Revision: DRAFT CORE DAMAGE ASSESSMENT GUIDANCE EPP 01-2.4

. . b Reference Use Page 9 of 13 4

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a. Obtain containment hydrogen concentration at 100% core over-temperature from table below 4

l Core Over-temperature Estimate Based on Containment Hydrogen Concentration RCS Pressure Water Injection Predicted Containment to the 7CS Dry Hydrogen Concentration

  • 8 Yes 7.4 volume percent 50 p No 3.8 volume percent Greater Than Yes 5.9 volume percent 1050 psig-No 3.7 volume percent
  • 'Use Dry Hydrogen Concentrations in above table with SAM CA-01 Figure 3-la to obtain Wet Hydrogen Concentrations.
b. Obtain current containment hydrogen concentration
c. Estimate over-temperature damage using:

Current Hz Concentration

% Core Damage" =

Predicted H2 Concentration at 100% Overtemp Damage

d. Containment radiation monitor and core exit thermocouple estimates should not deviate from hydrogen estimate by more than 25% damage.
2. If expected response is not obtained, determine if the deviation can be explained from the' accident progression.
  • Injection of vate; to the RCS
  • Bleed paths from the RCS e Direct radiation to the containment radiation monitors e Hydrogen burn in containment or hydrogen ignition source

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Revi'sion: DRAFT CORE DESAGE ASSESSMENT GUIDANCE EPP 01-2.4 w

Reference Use Page 10 of 13 7.3.4 Report clad damage er,timate to the Technical Support L Center Engineering Coordinator.

1. If clad damage estimates have increased by more than 1% in the past 30 minutes or if estimates exceed 5% clad damage, report possible change in Emergency Action Level to the Technical Support Center Engineering Coordinator.

7.3.5 Return to Step 7.1 in Diagnostic Section.

8.0 INITIAL ACTIONS 8.1 None 9.0 SUBSEQUENT ACTIONS 9.1 Report Findings 9.1.1 If clad damage estimates have increased by more than 1% in the past 30 minutes or if estimates exceed 5%

clad damage, report possible change in Emergency Action Level to the Technical Support Center Engineering Coordinator.

9.1.2 Report clad damage estimate to the Technical Support Center Engineering Coordinator.

9.2 Return to Step 7.1 in Diagnostic Section.

10.0 RECORDS 10.1 None 11.0 FORMS 11.1 None

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Revision: DRAFT CORE DAMAGE ASSESSMENT GUIDANCE EPP 01-2.4 w

Reference Use Page 11 c. .3 FIGURE 1 CONTAINMENT DOSE RATE RESPONSE TO TECHNICAL SPECIFICATION 3.4.8 SOURCE TERM WITH PRE-EXISTING IODINE SPIKE Containment Radiation Level vs. Time 10% Technical Specification 3.4.8 Reactor Coolant System Activity Source Term 2.9 2.8 - -

RCS Activity @ 10% of 100/E-Bar Including Pre-Existing Iodine Spike of 60 uCi/gm of I-131

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0 2.5 2.4 - - -

2.3 , , , ,

0 10 20 30 40 50 Time Since Shutdown (Hrs)

3 + ,, Revision: DRAFT CORE DAMAGE' ASSESSMENT GUIDANCE EPP 01-2.4 l

Reference Use Page 12 of 13 i

FIGURE 2 CONTAINMENT-DOSE RATE RESPONSE'TO GAP RELEASE SOURCE TERM i

l Containment Radiation Level vs. Time I Gap Release Source Term I

I- RCS Pressure < 1600 psig w/o Cont. Sprays 1E+4 - -

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RCS Pressure < 1600 psig w/ Cont. Sprays

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RCS Pressure > 1600 psig w/o Cont. Sprays _

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l RCS Pressure > l600 ps ig w/ Cont. Sprays IE+1 , , , .

0 10 20 30 40 50 l

Time Since Shutdown (Hrs)

-+-RCS Pressure < 1600 psig w/o Sprays -e-RCS Pressure < 1600 psig w/ Sprays 1 l 1 RCS Pressure > 1600 psig w/o Sprays -M-RCS Piessure > 1600 psig w/ Sprays j 1

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  • Revision: DRAFT CORE DAMAGE ASSESSMENT GUIDANCE EPP 01-2.4 Reference Use Page 13 of 13 I i 1 J

FIGURE 3 CONTAINMENT DOSE RATE RESPONSE TO CORE OVERTEMPERATURE SOURCE TERM Con +ainment Radiation Level vs. Time Core Over-temperature Source Term RCS Pressure < 1600 psig w/o Cont.

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RCS Pressure < 1600 psig w/ Cont. Sprays  ;

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f[ p For High Level Core Damdi Assessment Reduce by Factor of 100 k5 IE+4 - '

5 (R/lIr
  • 0.01) l a

f RCS Pressure > 1600 psig w/o Cont. Sprays  %

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x

( RCS Pressure > 1600 psig w/ Con:. Sprays IE+2 , , , ,

0 10 20 30 40 50 Time Since Stutdown (Hrs) l

+RCS Pressure < 1600 psig w/o Sprays -m-RCS Pressure < 1600 psig w/ Sprays

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-*-RCS Pressure > 1600 psig w/o Sprays -M-RCS Pressure > 1600 psig w/ Snisp  !

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