ML032760146
| ML032760146 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Clinton, Quad Cities, LaSalle  |
| Issue date: | 09/19/2003 |
| From: | AmerGen Energy Co, Exelon Generation Co, Exelon Nuclear |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| RS-03-186 | |
| Download: ML032760146 (35) | |
Text
ATTACHMENT A EP-AA-110-301, CORE DAMAGE ASSESSMENT (BWR)
EP-AA-1 10-301 Exeken.
Revision I Page 1 of 34 Nuclear Level 2-Reference Use CORE DAMAGE ASSESSMENT (BWR)
- 1.
PURPOSE 1.1.
The purpose of this attachment is to provide emergency response personnel with the methodology to estimate the degree of possible core damage at Exelon Nuclear's Boiling Water Reactor (BWR) stations. REFER to EP-AA-110-302 for methodology to estimate potential core damage for a Pressurized Water Reactor (PWR).
1.2.
This Core Damage Assessment process is designed to assist in estimating core damage after an accident with potential clad or core damage conditions.
This is done to assist in:
1.2.1.
Determining if the fuel barriers are breached to evaluate the appropriate Emergency Action Level (EAL) classification.
1.2.2.
Providing input on core configuration (coolable or uncoolable) for prioritization of mitigating activities.
1.2.3.
Determining the potential quantity and isotopic mix of a radiological release to project offsite doses.
1.2.4.
Predicting the radiation protection actions that should be considered for long term recovery activities.
1.2.5.
Satisfying inquiries from local and federal government agencies and provide evidence that the utility knows the plant conditions.
1.3.
Core damage may be assessed by:
Evaluating the drywell radiation levels (and confirmed by evaluating the extent of time the core was uncovered),
Concentration of certain isotopes in a reactor coolant analysis, or Concentration of hydrogen in the primary containment.
History of Core Cooling This procedure is intended to provide an acceptable alternative to existing station core damage assessment models and methods utilized by Reactor Engineering.
EP-AA-1 10-301 Revision 1 Page 2 of 34
- 2.
TERMS AND DEFINITIONS 2.1.
BWR-Boiling Water Reactor 2.2.
Cladding - The outer coating (usually zirconium alloy, aluminum or stainless steel), which covers the nuclear fuel elements to prevent corrosion of the fuel and the release of fission products into the coolant.
2.3.
Containment Type -
Clinton (Mark ll)
Dresden (Mark I)
LaSalle (Mark II)
Limerick (Mark II): 764 assemblies Containment Volume (384,570 ft3) = SuppressIon Pool (149,380
)
Drywell (235, 190 fe)
Peach Bottom (Mark ): 764 assemblies Containment Volume (303,600 ft3) = Suppression Pool (127,800 f) + Drywell (175, 800 ft3)
Quad Cities (Mark I) 2.4.
Core Release Fraction - The fraction of each isotope in the core inventory that is assumed to be released from the core under given core conditions.
2.5.
Cladding Failure
- 1.
Also referred to as "Cladding Oxidation", "Gap Release' or "Clad Rupture" in other documents.
- 2.
100% clad failure refers to the rupture of 100% of the fuel rods in the core. This would result in all fission products contained in the gap space being released to the reactor coolant system.
2.6.
Fission Products - The nuclei (fission fragments) formed by the fission of heavy elements or by subsequent radioactive decay of the fission fragments.
2.7.
Fuel Melt
- 1.
Referred to as "Core Melt" "In-Vessel Melt" or "Over-temperature" damage in reference documents.
- 2.
100% fuel melt refers to high temperatures in the fuel pellets in 100%
of the fuel rods in the core. This would result in all the fission products contained in the fuel pellet matrix being released to the reactor coolant system.
2.8.
Gap - The space inside a reactor fuel rod that exists between the fuel pellet and the fuel rod cladding.
EP-AA-1 10-301 Revision 1 Page of 34 2.9.
Gap Release - The release into containment of fission products in the fuel pin gap.
2.10. In-Vessel Core Melt - A condition during a reactor accident in which some of the cladding or reactor fuel melts as a result of overheating the fuel and remains inside the reactor vessel.
2.11. In-Vessel Core Melt Release - A release into containment from the reactor vessel, which assumes the entire core has melted, releasing a representative mixture of radioisotopes.
2.12. Minimum Steam Cooling RPV Water Level (MSCRWL) - The lowest RP water level at which the covered portion of the reactor core will generate sufficient steam to maintain the hottest clad temperature below 1 500'F.
2.13. Minimum Zero-Iniection RPV Water Level (MZIRWL) - The lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to maintain the hottest clad temperature below 1 8000F, assuming no injection into the RPV.
2.14. Shutdown -
As defined by station emergency operating procedures, normally assumed to be less than 4% for a BWR.
2.15. Slump - Relocation of molten reactor core during an accident.
2.16. Source Term - The amount and isotopic composition of material released or the release rate, used in modeling releases of material to the environment.
2.17. Spiked Coolant - Reactor coolant containing increased concentrations of non-noble isotopes, sometimes seen with rapid shutdown or depressurization of primary system.
2.18. Siked Coolant Release - The release into containment of 100 times the non-noble gas fission products found in the coolant.
2.19. Subcritical - The reactor condition when the number of neutrons released by the fission is not sufficient to achieve a self-sustaining nuclear chain reaction.
Defined under station emergency operating procedures.
2.20. TID - Total Isotopic Distribution 2.21. Vessel Melt-Throuah
- 1.
Referred to as "Ex-Vessel Melt" or"Melt Release" in reference documents.
- 2.
Core debris is relocated to the primary containment building after the reactor pressure vessel has failed.
EP-AA-110-301 Revision 1 Page 4 of 34
- 3.
RESPONSIBILITIES 3.1.
The TSC Core/Thermal Hydraulic Engineer shall serve as the Core Damage Assessment Methodology (CDAM) Evaluator.
3.2.
The TSC Radiation Controls Engineer shall coordinate radiological and chemistry information with the Core/Thermal Hydraulic Engineer in support of core damage assessment.
3.3.
The TSC Manager shall coordinate core damage assessment activities.
- 4.
MAIN BODY 4.1.
Use the appropriate Attachment to assess status of the affected unit's reactor core:
4.1.1.
For Mid Alantic ROG BWR Stations:
- 1.
ASSESS the status of critical BWR safety functions and MONITOR for indications that the core may already been uncovered or may soon become uncovered.
- 2.
PROJECT core damage if uncovered using the following methods, as applicable:
-, "Core Uncovery Time Method"
-, Containment Hydrogen Concentration Method"
- 3.
MONITOR radiation levels in containment to confirm and assess core damage.
- 4.
PROJECT core damage using Containment Radiation Level Method (Attachment 1).
- 5.
CONTINUE to assess core damage a reactor coolant sample to confirm extent of core damage using the Attachment 2 (Reactor Coolant Sampling Method).
- 6.
CONTINUE to assess core damage and INFORM those assessing consequences, classification and protective actions.
4.1.2.
For Mid West ROG BWR Stations refer to Attachment 5, BWR CDAM User Guide, to use the Core Damage Assessment Software Program.
I
EP-AA-1 10-301 Revision 1 Page b of 34
- 5.
DOCUMENTATION 5.1.
Reactor core damage assessment methods derived from U.S. NRC RTM-96, Chapter A (Reactor Core Damage Assessment). Figures included in RTM-96, Chapter A were generated using CONDOS II (NUREG/CR-2068).
5.2.
A Summary Form and method specific reports are generated by the BWR CDAM Software for use in documenting the results of the assessment.
- 6.
REFERENCES 6.1.
U.S. Nuclear Regulatory Commission (NRC) Response Technical Manual (RTM) 96, March 1996 6.2.
NEDO-22214, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions 6.3.
NEDC-33045P, Rev 0 (July 2001), Methods of Estimating Core Damage in BWRs 6.4.
WCAP-14696 (July 1996) Westinghouse Owners Group Core Damage Assessment Guidance.
6.5.
WCAP-14696-A (November 1999), Westinghouse Owners Group Core Damage Assessment Guidance.
6.6.
NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Accidents' 6.7.
Exelon Nuclear Radiological Emergency Plan Annex for Limerick Generating Station 6.8.
Exelon Nuclear Radiological Emergency Plan Annex for Peach Bottom Atomic Power Station EP-MW-123-1003, Rev.0, Core Damage Assessment Methodology (CDAM) Program Technical Basis
- 7.
ATTACHMENTS 7.1., Containment Radiation Level Method 7.2., Reactor Coolant Sampling Method 7.3., Core Uncovery Time Method 7.4., Containment Hydrogen Concentration Method 7.5., BWR CDAM User's Guide
EP-AA-1 10-301 Revision 1 Page 6 of 34 ATTACHMENT I CONTAINMENT RADIATION LEVEL METHOD Page 1 of 8 Per RTM-96, Chapter A (Method A.4), these calculations should provide the maximum reading expected under the conditions stated. The calculations assume: (1) a prompt release to containment of all fission products in the coolant, spike, gap or from in-vessel core melt; (2) uniform mixing in the containment; and (3) an unshielded monitor that can see most of the containment volume (minus torus / suppression pool volume). Because the mix is most likely different from that assumed in the calibration of the monitor, the actual reading at the upper end of the scale could differ by a factor of 10-100 if a shielded detector is used for higher radiation measurements.
The levels of damage indicated should be considered minimum levels unless there are inconsistent monitor readings. Inconsistent readings may be caused by the uneven mixing in containment (e.g., steam rising to top of dome, not enough time for uniform mixing to occur).
Figures have been prepared for each station, which represent plots of percent core damage versus containment radiation readings and percent clad damage versus containment radiation readings. Uncorrected containment radiation readings shall be used to provide a preliminary estimate of percent core and clad damage during the first ten (10) hours following a reactor shutdown.
CAUTION Containment radiation levels cannot be used to assess core damage in all cases. The release may bypass the containment, be retained in the coolant, be released over a long period of time, or not be uniformly mixed. Therefore, a low containment radiation level reading does not guarantee a lack of core damage.
This method should not be used if containment radiation monitors do not "see" more than 50% of Drywell volume.
- 1. MONITOR containment radiation levels to assess the degree of clad and/or core damage.
NOTE:
The readings taken following a Loss of Coolant Accident (LOCA) into the primary containment will be most accurate using the containment radiation monitors.
- 2. DETERMINE the time following the event and the uncorrected containment radiation reading and compare to the following figures for assessing the degree of core or clad damage.
EP-AA-1 10-301 Revision 1 Page 7 of 34 ATTACHMENT I CONTAINMENT RADIATION LEVEL METHOD Page 2 of 8 Limerick Il______________:______
REFER to Figure 1-1 for PERCENT OF FUEL INVENTORY AIRBORNE-IN THE CONTAINMENT VERSUS APPROXIMATE SOURCE AND DAMAGE ESTIMATE I Peach Bottom REFER to Figure 1-2 for PERCENT OF FUEL INVENTORY AIRBORNE IN I
THE CONTAINMENT VERSUS APPROXIMATE SOURCE AND DAMAGE l
ESTIMATE l
a
EP-AA-1 10-301 Revision I Page 8 of 34 ATTACHMENT I CONTAINMENT RADIATION LEVEL METHOD Page 3 of 8 FIGURE 1-1 LIMERICK GENERATING STATION Percent of Fuel Inventory Airborne in the Containment vs. Approximate Source and Damage Estimate Page 1 of 3
% Fuel Curve Inventory No.
Released Approximate Source and Damage Estimate 1
100.
100% TID-14844, 100% fuel damage, potential core met.
- 50.
50% TID noble gases, TMI source.
2
- 10.
10% TID, 100% NRC gap activity, total clad failure, partial core uncovered.
- 3.
3% TID, 100% WASH-1400 gap activity, major clad failure.
z 3
- 1.
1% TID, 10% NRC gap, Max. 10% clad failure.
0 4
.1
.1% TID, 10% NRC gap, 1% clad failure, local heating of 5-10 fuel assemblies.
5
.01
.01% TID,.1% NRC gap, clad failure of 3/4 of a fuel o
assembly (36 rods).
6 10-4
.01% NRC gap, clad failure of a few rods a
10 4 100% coolant release with spiking.
7 5X104 100% coolant inventory release.
106 Upper range of normal airborne noble gas activity in containment.
100% Fuel Inventory = 100% Noble Gases +25% Iodine + 1% particulates.
NOTES (1) These curves account for the finite containment volume and shield wall seen by the detector but do not account for any monitor physical or shielding characteristics or calibration uncertainties.
(2) The curves assume that both airborne noble gases and iodines are significant Sprays (if used) would make the iodine and particulate contribution (presently about 50%) insignificant However, particulate plateout on the monitor casing and direct shine doses from components may make the readings unreliable.
(3) Curve uncertainties are on the order of a factor of 5 to 10.
EP-AA-1 10-301 Revision I Page 9 of 34 ATTACHMENT 1 CONTAINMENT RADIATION LEVEL METHOD Page 4 of 8 FIGURE 1-1 LIMERICK GENERATING STATION Percent of Fuel Inventory Airborne in the Containment vs. Approximate Source and Damage Estimate Page2 fO3 LGS CONTAINMENT RADIATION MONITOR CURVES I
0e a
I L)
I I 07 -
0 4 Lao
_______________ ~
~
~
z 0
CD) 0d rJ 1
l l
10-,
10° 101 Ilo 101 TIME AFTER SHUTDOWN (HRS) 1 - 100%
2 - 10%
3 - 1%
4 - 0.1%
5 - 0.01%
6 - 0.001%
EP-AA-1 10-301 Revision I Page 10 of 34 ATTACHMENT I CONTAINMENT RADIATION LEVEL METHOD Page 5of8 FIGURE 1-1 LIMERICK GENERATING STATION Percent of Fuel Inventory Airborne in the Containment vs. Approximate Source and Damage Estimate Page 3 of 3 LGS CONTAINMENT RADIATION MONITOR CURVES I
I 0
I R
102 la' 00 z
0 LU 10' 10 I
_l-1Go l'
102 io0 TIME AFTER SHUTDOWN (HRS) 4 - 0.1%
5 - 0.01%
6 - 0.001%
7-100%
TID TID TID REACTOR COOLANT
EP-AA-1 10-301 Revision I Page 11 of 34 ATTACHMENT I CONTAINMENT RADIATION LEVEL METHOD Page 6 of 8 FIGURE 1-2 PEACH BOTTOM ATOMIC POWER STATION Percent of Fuel Inventory Airborne in the Containment vs. Approximate Source and Damage Estimate Page 1 of 3
% Fuel Curve Inventory No.
Released Approximate Source and Damage Estimate 100.
100% TID-14844, 100% fuel damage, potential core melt.
- 50.
50% TID noble gases, TMI source.
2
- 10.
10% TID, 100% NRC gap activity, total clad failure, partial core uncovered.
- 3.
3% TID, 100% WASH-1400 gap activity, major clad z
failure.
3
- 1.
1% TID, 10% NRC gap, Max. 10% clad failure.
4
.1
.1% TID, 10% NRC gap, 1% clad failure, local heating of S
5-10 fuel assemblies.
0 I
5
.01
.01% TID,.1% NRC gap, clad failure of 3/4 of a fuel U
assembly (36 rods).
6 104
.01% NRC gap, clad failure of a few rods 104 100% coolant release with spiking.
7 5X104 100% coolant inventory release.
I 04 Upper range of normal airborne noble gas activity in containment.
100% Fuel Inventory = 100% Noble Gases +25% Iodine + 1% particulates.
NOTES (1) These curves account for the finite containment volume and shield wall seen by the detector but do not account for any monitor physical or shielding characteristics or calibration uncertainties.
(2) The curves assume that both airborne noble gases and iodines are significant Sprays (if used) would make the iodine and particulate contribution (presently about 50%) insignificant However, particulate plateout on the monitor casing and direct shine doses from components may make the readings unreliable.
(3)Curve uncertainties are on the order of a factor of 5 to 10.
EP-AA-1 10-301 Revision 1 Page 12 of 34 ATTACHMENT I CONTAINMENT RADIATION LEVEL METHOD Page 7 of 8 FIGURE 1-2 PEACH BOTTOM ATOMIC POWER STATION Percent of Fuel Inventory Airborne in the Containment vs. Approximate Source and Damage Estimate Page 2 of 3 PEACH BOTTOM CONTAINMENT RADIATION MONITOR CURVES I
0 1 log M
'a 0
i 0
104 103 102 lO, z
0 P
C-t0 0
C.)
wU
(.
I On I 0-.
too log 102 TIME AFTER SHUTDOWN (HRS) 1o3 PEACE BOTTOM MONITOR RESPONSE CURVES rOR PRIMARY CONTAINMENT LOW RANGZ MONITORS CURVE INDEX
- 1.
100%
rUZL INVENTORY (100% TID-14844)
- 2.
10 FUZL INVENTORY (100% GAP ACTIVITY/R.G. 1.25)
- 3.
1 FUEL INVENTORY (10% NRC GAP-CLAD FAILURE)
- 4.
0.1%
FUZL INVENTORY
- 5.
0.01%
rUEL INVENTORY
- 6.
0.001% rUEL INVENTORY
EP-AA-1 10-301 Revision I Page 13 of 34 ATTACHMENT I CONTAINMENT RADIATION LEVEL METHOD Page 8 of 8 FIGURE 1-2 PEACH BOTTOM ATOMIC POWER STATION Percent of Fuel Inventory Airborne in the Containment vs. Approximate Source and Damage Estimate Page 3 of 3 PEACH BOTTOM CONTAINMENT RADIATION MONITOR CURVES I
1 o2 I
I z
0 Co 0
0 0
TIME AFTER SHUTDOWN (HRS)
PEACH BOTTOM MONITOR RESPONSE CURVES FOR PRIMARY CONTAINMENT LOW RANGE MONITORS CURVE INDEX 4 -
0.1%
FUEL INVENTORY 5 -
.01%
FUEL INVENTORY 6 -
.001% FUEL INVENTORY 7 -
100%
COOLANT ACTIVITY
EP-AA-1 10-301 Revision 1 Page 14 of 34 A1TACHMENT 2 REACTOR COOLANT SAMPLING METHOD Page 1 of I Per RTM-96, Chapter A (Method A.5), for BWR this method of confirming core damage assumes that releases from the core are uniformly mixed in the coolant and torus I suppression pool with there is no dilution from injection. If most of the core release is confined to the reactor coolant system, the concentrations in the coolant could be up to 10 times higher. The baseline coolant concentrations are for 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after shutdown of a core that has been through at least one refueling cycle. The half-life of the fission products should be considered in analyzing samples. The plant-specific coolant system volume does not have a major influence on coolant concentrations (< 20%).
CAUTION Since samples may require hours to collect and analyze, and may not be representative of primary system concentrations (e.g., no flow through sample line), coolant concentrations should NOT be required to confirm core damage.
- 1. COMPARE coolant isotopic concentrations with the concentrations listed in Table 2-1 to estimate core damage for gap releases or in-vessel melting.
Table 2-1 BASELINE BWR COOLANT CONCENTRATION' Normal Concentration Concentration after Nuclide Concentration2 after gap release3 in-vessel melt3 (uCi/g)
(uCilg)
(uCi/g) 131 2 x 10-3 1 x 103 1 x 104 133 IV 1x0-2 3x103 2x104 135i 0
2x 10-2 2x 103 2x104 134Cs 3 x 10-5 x 10 2 6 x 10 2 137Cs 8x 10-5 8 x 10 4 x 102
_ 4_Ba NC NC 2 x 103 90Sr 7 x 106 NC 1 x 103 NC - Not Calculated 1 Table 2-1 is consistent with Table A-4 (BWR Baseline Coolant Concentration) from U.S. NRC Response Technical Manual (RTM) 96, March 1996 2 Source for Onormal coolant' is ANSI/ANS 18.1, 1984 (confirmed by NUREG/CR-4245, Table 3.2) 3 Concentration in the reactor coolant system and torus / suppression pool
EP-AA-1 10-301 Revision 1 Page 15 of 34 ATTACHMENT 3 CORE UNCOVERY TIME METHOD Page of I Per RTM-96, Chapter A (Method A.3), it can be assumed that the fuel in the core will heat up at 1-20F per second within 5-10 minutes after the top of an active core isuncovered. These heatup estimates are reasonable within a factor of two if the core is uncovered within a few hours of shutdown (including failure to scram) for a boil-down case (without injection). If there is injection, core heatup may be stopped or slowed because of steam cooling. However, steam cooling may not prevent core damage under accident conditions.
- 1. DETERMINE the time that the core was or is projected to be uncovered.
- 2. ESTIMATE the possible core damage by comparing the time of core uncovery to the values in Table 3-1.
Table 3-1 CORE DAMAGE VS. TIME THAT REACTOR CORE IS UNCOVERED4 lime 20% of the Core Core core is uncovered Temperature Temperature Possible Core Damage (minutes)
(OF)
(C) 0
>600
>315 None 30-45 1800-2400 980-1300 GAP Release Local fuel melting Burning of cladding with steam production (exothermic Zr-H20 reaction with rapid H2 generation)
Rapid fuel cladding failure (gap release core) 30-90 2400-4200 1300-2300 In-Vessel Melt Rapid release of volatile fission products (in-vessel severe damage release from the core)
Possible relocation (slump) of molten core Possible uncoolable core 120-180+
>4200
>2300 Vessel Melt-Through Melt-through of vessel with possible containment failure and release of additional less-volatile fission products 4 Table 3-1 is consistent with Table A-2 (Core Damage vs. Time That Reactor Core Is Covered) from U.S. NRC Response Technical Manual (RTM) 96, March 1996
EP-AA-1 10-301 Revision 1 Page 16 of 34 ATTACHMENT 4 CONTAINMENT HYDROGEN CONCENTRATION METHOD Page 1 of I This method may be used to assess the core damage based on the hydrogen concentration in samples of the containment atmosphere.
Per RTM-96, Chapter A (Method A.6), the hvdro-ien concentrations using this method are for wet samples; however, most hydrogen samples are dry (steam removed). If a dry sample concentration is used, one may overestimate considerably the level of core damage. This method assumes that all hydrogen is released to the containment and is completely mixed in the containment atmosphere.
CAUTION Since containment samples may require hours to collect and analyze, and may not be representative of actual concentrations, hydrogen should NOT be relied upon to confirm core damage in all cases.
- 1. COMPARE the hydrogen percentage obtained to the graph in Figure 3 to estimate percentage of metal-water reaction and possible levels of core damage.
% MetAWter Reacton & Core Darnage tate s0 40 30 20 46 Possibk Melt mgh 4.Ps+ l Unixlbl core
<. str Fuel Melt Ad F*IUe 0
0.1 I
10 H2 % In Containment 100 EBWRMk 1i
- -BWR Mk I &
Figure 4-1s I
5 Figure 4-1 is consistent with Figure A-13 (Percentage of Hydrogen in Containment Relative to Core Damage) from U.S. NRC Response Technical Manual (RTM) 96, March 1996
EP-AA-I 10-301 Revision 1 Page 17 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 1 OF 18
- 1.
OVERVIEW 1.1.
As a windows based application designed in access, bwr cdam, uses many standard user interfaces. Instructions are not provided in basic computer operations in the windows%) environment. The user must be familiar with these to efficiently operate the program.
1.2.
It is also assumed user is familiar with basic reactor physics and core damage fundamentals. Emergency Response Organization training will provide an overview of core damage assessment methodologies.
1.3.
The program should be used by qualified personnel as a tool to estimate the type and amount of core damage.
- 2.
DETERMINE APPROPRIATE AND AVAILABLE ASSESSMENT METHODS lMd-West ROGl lREFER to EP-MW-1 1 0-1 001 for a listing of appropriate plant parameter l
lpoints to be used following a LOCA.
2.1.
The magnitude and type of event, transport mechanism and time after shutdown will be influencing factors on the method(s) utilized to determine the extent of core damage. Damage estimates can be developed using one or more methods as they become available or applicable.
2.1.1. Indications Of Core Damage A.
The primary indicators of core damage that are available during the early phases of an event:
B.
Drywell/Containment Radiation Monitor Readings C.
Drywell/Containment Hydrogen Readings 2.1.2. Auxiliary indicators that are used to confirm and better define the possible type of damage are:
A.
Reactor Pressure Vessel Level Indication System readings B.
Estimation of maximum temperature reached within the core C.
Estimated core uncovery time D.
Abnormal Source Range Monitor readings
EP-AA-1 1 0-301 Revision I Page 18 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 2 OF 18 2.1.3. Long Term Indicators (once liquid or gaseous samples can be safely obtained) are:
A.
Isotopic Ratios B.
Presence of high levels of rare isotopes C.
Quantity of isotopes present in samples 2.2.
Choose the assessment method(s) most appropriate for the existing conditions.
Methods available for assisting in the determination of the extent of core damage include the following:
Method Use Comment Containment Early Indication of Core Uncertainties due to variables in release Radiation Monitor Damage of fission products from RCS and effects of containment sprays.
Core Conditions Indication of onset of May not be reliable during later phases of Core Damage core overheating due to changes in core geometry.
RPV Level Indication of Core Indicates possible damage not useful in Uncovery estimating the quantity of damage.
Source Range Indication of Core Loss of water level leads to increase in Monitor Uncovery gamma detection.
Containment Early Indication of Core Significant uncertainties due to variable Hydrogen Monitor Damage Hydrogen generation in core and in release of Hydrogen from RCS and effects of containment sprays.
RCS Samples and Late Indication of Core Very large uncertainties until all systems Containment Sump Damage-Suppression have reached equilibrium. Useful in and Atmosphere Pool Samples provide planning long term recovery.
Samples indication of Rx Vessel Failure
- 3.
START UP 3.1.
The application is accessed by one of the following:
3.1.1. Open the BWR CDAM desktop icon on applicable dose assessment computers.
A.
Start the BWR CDAM program for the plant that has declared an emergency.
B.
Programs are labeled BWR CDAM.
3.1.2. Select run from the 'start bar' and type in the file path and name as follows:
C:\\CDAM\\BWR CDAM.MDB
EP-AA-1 10-301 Revision 1 Page 19 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 3 OF 18 3.2.
IF the assigned Core Damage Assessment Computer cannot access the application or the CDAM program will not run, THEN Install BWR CDAM on any computer from CDs or Disks located in the TSC or the EOF Library. CDAM is installed by copying appropriate file to computer's hard drive.
- 4.
SELECTION AND PERFORMANCE OF ASSESSMENT 4.1.
Choose the assessment method(s) most appropriate for the existing conditions.
Methods available for assisting in the determination of the extent of core damage include the following:
4.1.1. Containment Radiation Analysis - (Section 5.2) 4.1.2. Core Conditions Analysis (Cooling History) - (Section 5.3) 4.1.3. Containment Hydrogen Analysis - (Section 5.4) 4.1.4. Nuclide Analyses (Ratios and Abnormal Isotopes) - (Section 5.5) 4.1.5. Liquid Samples Analysis - (Section 5.6) 4.1.6. Gaseous Samples Analysis - (Section 5.7)
EP-AA-1 10-301 Revision 1 Page 20 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 4 OF 18 4.2.
Basic Program Flow Diagram Summary Screen Contmnt core Contmnt Uid Liquid Gaseous Rad Conditions Hydrogen Analysis Sample Sa pe System Volumes Liquid Gaseous Graphs Graphs
- 5.
PROGRAM SCREENS AND INPUTS 5.1.
Main Screen - Summary Paae
I i
I EP-AA-1 10-301 Revision 1 Page 2 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 5 OF 18 5.1.1. When program is started the following screen appears: (boxes are empty when program is originally launched.)
"Mll 11Mr
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l CAUTION Selecting an "Affected Station" resets all inputs to default values.
5.1.2. SELECT the Affected Station before other "Assessment Methods".
CAUTION Pressing the "Quit" button exits the program. When the program is closed all data is reset. Program saves no information to disk; printed reports serve as record of core damage assessments.
5.2.
Containment Radiation Monitor Method 5.2.1. Pressing "Cont Rad Monitors" button opens the following form:
EP-AA-1 10-301 Revision 1 Page 22 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 6 OF 18 D tpa~Iamtet'et$,.';~>.i~tY>
- ^<I5
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l SS f
i~-
ii00,-'3.....................
Melt Ca 3 >CN05S E 29^
3 Damage~ Etimate:
4Z.Ad -l 7;0%l1i3 a
any.
060%Reading Rr 1
E 8 E03 i
W. R itig lRTi Note: The highet monitor iabemnt eltcalculaltinsg :g2\\
w'"
S.
i I"II! 1W_ ",
"'
1--
--
1-
11 Izl I I f 11 `
"-nl
r,
- 7 M>=elt-Chid Fltimated readIng within 05 1
~Conainmnt i usd fo thet Preliminary results (affect of input data) are shown here See 5.2.7 II II t3fa Al-IJ11jieset:
e4 44 B ack 44 5.2.2. Highest drywell or containment / Suppression Chamber radiation monitor reading that occurred is entered in these boxes. Program allow entry from 2 high range monitors for drywell location and 1 for Torus or Suppression Chamber, however a reading may be entered from any monitor which accurately showed containment radiation levels. If two entries are made only the highest is used.
5.2.3. Drywell/Containment Spray A. IF the Drywell/Containment Spray system was operated for the majority of the time since the estimated time of the onset of core damage THEN choose "Drywell Spray On".
B. IF the Drywell/Containment Spray system was NOT operated or only operated briefly (e.g., <10% of time since the estimated time of the onset of core damage) THEN choose "Drywell Spray Off'.
5.2.4. Enter the time after reactor shutdown, which corresponds the time the containment radiation reading was taken. Value must be between 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown, which corresponds to the time period in which this method is considered effective.
5.2.5. Pressing "Reset" button resets values on this form only.
EP-AA-1 10-301 Revision I Page 23 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 7 OF 18 5.2.6. Pressing "Containment Graph" or "Supp Chamber Graph" button displays a screen similar to the following:
Id a
of erl~hixst Ls~eX~fi
~
i~tnimrd Mnhn Redii~ :11
~ ~ ~
~ ~ ~ ~ ~~~~~~~~~~~~~--
=___
Tm AtrS'
~~1 E + 4 -
- i - - - - - -
-~~~~~~~~~~
= -
~----------
_ _ _ __4__
'*'i i'.q.w4
_C l,d!
l
_ 4)t A 5 9 d,
_ _ _ _ _ _ _ _ _ _ _'_ _ _ _.., '^AL':
A. Graph shows high and low containment radiation levels which correspond to 100% Melt or Clad or 1% Melt or Clad damage. A dot shows the last containment radiation level entered into the program for assessment.
B. Pressing "Print" button will print report of containment radiation method inputs and best estimate of damage.
5.2.7. Pressing "Back" button takes the user back to the summary screen.
EP-AA-1 10-301 Revision 1 Page 24 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 8 OF 18 5.3.
Core Conditions Methods 5.3.1. Pressing "Core Conditions" button opens the following form:
lSee 5.3.2 F.,iPrae no.§ a
'The cowf lis p Sll11-If;d 20 bu i _ooed by steam. a tempelaturetsler expeed to ai elow 15at Femain coee 2:;
g t:oB
'dmwit:'!g'
'-.$Mwrpdb y rrii damage is Doetem 5.3.2. Reactor Pressure Vessel (RPV) Water Leve5l3.
5.3.3 Source Ran e oitr reading 10Ates thont expcted chekes.ekReut The.4 coeBas emoin t
co romeN Rector dOmaeraos er rtapid Z
ter reineein erandamage ma u
hae!tde h
is G
leas the fel ie do to he eitsted t
cuncodamadewithoutect e
amosrycoln raddinr cails.
~~ S~ee 5..__
5.3.2. Reactor Pressure Vessel (RPV) Water Level A.
Enter the lowest recorded (or estimated) RPV Level (range 0 to -
350) and core spray flow at time of lowest reading.
5.3.3. Source Range Monitor.
A.
Review plant parameter h istory and if the SRM had indications of a reading 1 0 times those expected check yes.
5.3.4. Based on inputs from Reactor Operators, TSC Staff and other engineering personnel (including outside sources such as General Electric personnel) enter the estimated time the reactor core (20% of top of active core) was uncovered without steam or spray cooling reactor core.
5.3.5. Based on inputs from Reactor Operators, TSC Staff and other engineering personnel (including outside sources such as General Electric personnel) enter the estimated highest temperature reached in the reactor core.
5.3.6. Pressing "Reset" button resets values on this form only.
EP-AA-1 10-301 Revision 1 Page 25 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 9 OF 18 5.3.7. Pressing "Print" button prints report of inputs and results of core temperature methods of core damage assessment.
5.3.8. Pressing "Back" button takes the user back to the summary screen.
5.4.
Containment Hydrogen Evaluations CAUTION This CDAM assumes no ignitor operation. Ignitor use limits containment hydrogen concentration affecting the reliability of this method.
5.4.1. Pressing "Cont Hydrogen" button opens the following form:-
- ~
etal-WaiciH ea~i~onLSee5.4.5
'4 5.4.2. Enter highest Drywell and/or Suppression Chamber hydrogen level measured. Suppression chamber reading can only be entered if user selects "no" under Equilibrium.
5.4.3. System Equilibrium:
EP-AA-1 10-301 Revision 1 Page 26 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 10 OF 18 A.
Select "Yes" for equilbrium if containment and suppression chamber monitors read the same or only atmospheres are assumed equalized.
B.
Select "No" for equilbrium if containment and suppression chamber atmospheres are not in equilibrium or only Containmen H2 reading is available.
5.4.4. Pressing "Print" button prints report of inputs and results of core level methods of core damage assessment.
5.4.5. Pressing "Back" button takes the user back to the summary screen.
5.5.
Nuclide Analysis 5.5.1. Pressing "Nuclide Analysis" button opens the following form:
I I
I E
I I
I / ~ ~ ~
~
~
~
~
~~~,n i I
~~-e
- 55.
yLipjbe Isotp
-s 1140 Tine0Sutownia I,
inge"ft e
i U
L
'I eB::Aciit el LSmtl IC
-r rl a
nI C l1.-
r- - -
1 -
-J-
-,I Alkaline E~~i14
- ~ I-rI AiAH E-.
pee a..-
-l- -E a1. E 1.+
~~i
_ _ i _ _ _ _ Di;Af
-£ 2r7
.OOE 4~f~
.2
.2
~
V~~e.1~~1m
~
~
4 0.4 1R Jht l_______
- 0. 6, V i bC 62i ii en 46t n-s u
MIC i A 112 2.OI~I~~ ~ 012 1
is m
See 5.5.
U p 1.10JE+0i j"jKjL-
-j!
q v-j lW i -5 L34
- £.-
V-=
=
=
E~-
5.5.2. Enter the time since reactor shutdown when the sample was taken and sample results in uCi/cc.
EP-AA-1 10-301 Revision 1 Page 27 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 11 OF 18 5.5.3. If the ratio is greater than predicted melt ratio, melt damage is predicted if less than clad ratio, clad damage is predicted.
A.
Noble Gases are ratioed to Xe-133 B.
Halogens are ratioed to 1-131 5.5.4. IF abnormal levels of rare isotopes are present THEN check yes AND check which isotopes are present.
5.6.
Liquid Samples 5.6.1. Pressing "Liquid Samples" button opens the following form:
I I
5.6.2. Select appropriate isotope.
5.6.3. Select sample location. If samples are available from both locations select both.
5.6.4. Enter sample activity(s) and Time After SD that samples were taken. If sample was taken from only one location and systems are in equilibrium check yes for "Systems in Equilibrium".
5.6.5. Enter power history (past to present, i.e. oldest steady state history as record number) of core since last refueling. Shutdown times are entered as the number of days with Ave Power (%) set at 0.
EP-AA-1 10-301 Revision 1 Page 28 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 12 OF 18 A.
For short-lived isotopes power history should extend at least 30 days.
B.
For long-lived isotopes power history should extend at least 100 days, however the power history for the extent of the cycle is preferred.
C.
Variations in steady state power should be limited to +/- 20% within each operational period entered.
5.6.6. Once all data has been entered pressing the Calculate" button will display the % Damage Estimates.
5.6.7. Pressing Volumes" button displays the follow screen:
~
~
~
~~
4 47~~~~~~2E+091
~~~~~~~~~~~~~~See 5.6.7 A.
Program enters default RCS volume, which the user may change based on RPV Level. Readings at time of sample.
B.
Program enters default Suppression Chamber volume, which the user may change based on readings at time of sample.
C.
Program enters default Containment free air volume which user, may change based on conditions at time of sample. Unless there has been significant flooding of drywell this value will not change.
D.
Program enters default Suppression Chamber free air volume which user may change based on conditions at time of sample. If there has been a significant increase or decrease in the water level in the Suppression Pool or Torus then the free air volume will change.
E.
Pressing "Reset" button resets all volumes to default values.
EP-AA-1 10-301 Revision 1 Page 29 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE13OF18 F.
Pressing "Back" button takes the user back to the Liquid or Gaseous screen, which user used to call volume form.
5.6.8. Pressing "Graph" button displays the following screen:
e ~ ~
~
~
~
IpIt4,byip
~~~=O-RN_
_a pl8 Lcso 10 C~eki IO I
I A.
Graph shows High, Low, and Best melt curves; High, Low, and Best clad damage curves, and a red line across graph indicating entered corrected sample activity.
B.
User can select "Print" button to print graph and summary of inputs or press "Back" button to go back to liquid or gaseous form which called this form.
5.6.9. Pressing "Back" button takes the user back to the summary screen.
EP-AA-1 10-301 Revision 1 Page 30 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE14OF18 5.7.
Gaseous Samples 5.7.1. Pressing "Gas Samples" button opens the following form:
Sanwfe S tiepng j4 2.OOE.OO f
- r. ;
- HiheV f e
a,7it~~~~~~
If111 111 v
10
~~~~~~~~~~~~~~~~See 5.7.4 1
~~~~~~~~~~.0+1, 5.7.2. Select appropriate isotope and sample location.
5.7.3. Enter Sample Information:
A.
Enter sample activity for selected isotope.
B.
Enter Time After SID that sample was taken.
C.
Enter the pressure and temperature of the system sampled D.
Enter the end pressure and temperature of sample.
5.7.4. Enter power history history (past to present, i.e. oldest steady state history as record number) of core since last refueling. Shutdown times are entered as the number of days with Ave Power (%) set at 0.
A.
For short-lived isotopes power history should extend at least 30 days.
EP-AA-1 10-301 Revision 1 Page 31 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 15 OF 18 B.
For long-lived isotopes power history should extend at least 100 days, however the power history for the extent of the cycle is preferred.
C.
Variations in steady state power should be limited to +/- 20% within each operational period entered.
5.7.5. Once all data has been entered pressing the "Calculate" button will display the % Damage Estimates.
5.7.6. Pressing "Volumes" button displays the follow screen (same as 5.6.7):
based-on RPV @Lrevel Readi at time of saEpe.
B.
Program enters default S
volume, which t he user may change based on readings at time of sample.
C.
Program enters default Containment free air volume which user may change based on conditions at time of sample. Unless there has been significant flooding of drywell this value will not change.
- 0.
Program enters default Suppression Chamber free air volume which user may change based on conditions'at time of sample. If there has been a significant increase or decrease in the water level in the Suppression Pool or Torus then the free air volume will change.
E.
Pressing "Reset" button resets all volumes to default values.
F.
Pressing "Back" button takes the user back to the Liquid or Gaseous screen, which user used to call volume form.
.D
EP-AA-1 10-301 Revision 1 Page 32 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 16 OF 18 5.7.7. Pressing "Graph" button displays the following screen:
I S I~~~mP~
am
_ __U_
1 0
'A M b__
N ~
~
~_
-7w~ i~T
!V 7_
m ~
V o n it ~~~~~~~~
A ~ ri~~*~:1 Sj~tom~~qui77U7 d
%C~~FeR_ 1000t~q x', rumjvL+,t~;+9 I
A.
Graph shows High, Low, and Best melt curves; High, Low, and Best clad damage curves, and a red line across graph indicating entered.
B.
User can select "Print" button to print graph and summary of inputs or press "Back" button to go back to liquid or gaseous form which called this form.
5.7.8. Pressing "Back" button takes the user back to the summary screen.
- 6.
CORE DAMAGE
SUMMARY
REPORT 6.1.
Once the program user enters data for all available assessment methods and the program calculates damage based on inputs, SELECT the "Print" button to print a summary of all methods used 6.2.
A sample report is shown on the next page.
6.3.
Individual tasked with assessing core damage shall then analyze report to determine best estimate of type and amount of damage.
EP-AA-1 10-301 Revision 1 Page 33 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE17OF 18 NOTE The CDAM program does not use the Fuel Overheat Condition Category 6.4.
Based on estimated type and amount of damage and following table (table also printed on summary report) assign NRC Core Condition Category (1-10).
Degree of Minor Intermediate Major Degradation
(<10%)
(10% to 50%)
(>50)
No Core Damage 1
1 I
Cladding Failure 2
3 4
Fuel Overheat 5
6 7
Fuel Melt i
10
- 7.
QUITING, OR EXITING, THE PROGRAM 7.1.
Pressing the "Quit" button on the Summary Screen exits the program.
7.1.1. When the program is closed all data is reset.
7.1.2. Program saves no information to disk; printed reports serve as record of core damage assessments.
EP-AA-1 10-301 Revision 1 Page 34 of 34 ATTACHMENT 5 BWR CDAM User's Guide PAGE 18 OF 18 SAMPLE
SUMMARY
REPORT CDAM Methoc Core Damage Summary o Drrsden 0 LaSafle a Quad ClUes Station: o Clinton ssessment Methods:
Melt Clad Contalnment Radiation Monlors Containment [
l Supperssion Chamber 1
l
-Il Core Conditions Core Cooling:
Clad Damage Core Uncovery Tie: l No Core Dnage l
SRM Count Rate: l No Core Dunage Core Terrp:
Clad Fallure Containment Hydrogen*
l l
l 2 8l 8ample Analysis Ratis: [
Fuel Melt Abnormal Isotopes: [
6 or 19 Present RCS: Liquid Samples: l0%
l l 0 l
Chamber Gas Samples: E g l
[1] 1l
- These mehods should NOT be used for quallite or quanttatle assasment except In he oase of a LOCA Anayst's Estimat:
o3 No Core Damage 0 Cladding Failure a Fuel Melt Amount NRC Core Condtion Category:
Degree of J Minor Itermediate Major Degradation
(<lav)
(101-50%)
(>SM)
Noa Core Darnage 1
1 1
Claddina Failure 2
3 4
Fuel Overheat 5
6 7
Fuel Mlelt 8
9 I10 Generated Bu:
Name:
Date:
120502 lime: