ML20137G100
| ML20137G100 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 08/23/1985 |
| From: | Bailey J, Bockhold G GEORGIA POWER CO. |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.B.3, TASK-TM GN-692, PROC-850823, NUDOCS 8508270175 | |
| Download: ML20137G100 (71) | |
Text
--.
,1 _a Appra Proczdura No.
Vogti] Electric Generating Pl;nt A NUCLEAR OPERATIONS 91502-C Revision No.
[ g 2/hf /
Unit COMMON Georgia Power e oe No.
1 of 46 c
C/::
CORE DAMAGE ASSESSMENT 1.0 PURPOSE This assify and estim. ough core fission product re ease vessel level indications, and core exit thermocouple temperatures together with additional auxiliary indicators.
2.0 PRECAUTIONS AND LIMITATIONS Exercise extreme caution during post accident sampling and analysis activities. Possible high dose rates at sampling stations and in laboratories require that appropriate radiological precautions be taken to ensure
. personnel safety. -
3.0 PREREQUISITES An emergency condition has been declared and' core damage is suspected, and the Emergency Director has directed implementation of this procedure.
4.0 RESPONSIBILITIES 4.1 CHEMISTRY DEPARTMENT 4.1.1 Chemistry personnel are responsible for overall coordination of this procedure including the assignment of responsibilities to other groups or individuals as is required to complete the assessment.
4.1.2 Chemistry personnel, as coordinators of assessment activities, are responsible for data management.
4.1.3 Chemistry personnel are responsible for making all damage estimates.
4.1.4 Chemistry personnel are responsible for post accident sampling and analysis activities and transmitting data to personnel coordinating core damage assessment activities.
8508270175 850823 PDR ADOCK 05000424 E PDR CONTINUED FC31Ct A R3' 30
. t PROCEDURE No. REVISloN PAGE No.
91502-C 0 2 of 46 4.2 OPERATIONS DEPARTMENT 4.2.1 Operations Department personnel are responsible for collecting, recording and transmitting to the Chemistry personnel coordinating damage assessment activities, data provided by Control Room instrumentation through implementation of Procedure 91503-C, " Control Room Instrumentation Output For Assessment Of Core Damage".
4.3 PLANT ENGINEERING DEPARTMENT 4.3.1 Engineering personnel are responsible for implementing Procedure 91504-C, " Core Inventory Determinations Using Reactor Power History", and for transmitting the completed results to Chemistry personnel responsible for coordinating damage assessment activities.
5.0 MAIN BODY 5.1 EVALUATION OF INITIATING EVENT AND PRELIMINARY INDICATIONS OF CORE INVENTORY NOTES
- ~
- a. No generalized core damage is
_ likely-if the fuel assemblies have not been uncovered. If the RVLIS full range instrument indicates that the collapsed liquid level has never been below the top of the core and no core exit thermocouple temperatures corresponding to superheated steam at the corresponding RCS pressure were indicated, then no generalized core damage is probable,
- b. If the core was uncovered, increases in the CVCS letdown ~
monitor, containment atmosphere process radiation monitor, and the Containment Building area radiation monitors are evidence that some degree of fuel damage occurred.
CONTINUED rwn
m -
a f PROCEDURE No. REVISloN PAGE NO.
91502-C 0 3 of 46 5.1.1 Chemistry personnel requests the TSC Manager to have the Engineering group implement Procedure 91504-C,
" Reactor Power History And Core Inventory Determinations" and deliver completed data sheets to the Chemistry personnel coordinating damage assessment activities.
5.1.2 Chemistry personnel requests the Operations Department to implement Procedure 91503-C, " Control Room Instrumentation Output For Assessment of Core Damage" and deliver completed data sheets to the Chemistry personnel coordinating damage assessment activities.
5.1.3 When obtained, compare the thermocouple temperatures recorded on Data Sheet 1 of Procedure 91503-C, " Control Room Instrumentation Output For Assessment Of Core Damage" to those in Table 1, " Fuel Damage Versus Elevated Fuel Rod Temperatures".
5.1.4 Record the maxinum thermocouple temperature obse'rved and the corresponding damage category from Table 1, on Data Sheet 1, Sheet 1.
5.1.5 Record the RVLIS indication on Data Sheet 1, Sheet 1.
5.1.6 Obtain the containment high range area monitor reading from-Procedure 91503-C, " Control Room Instrumentation Output For Assessment Of Core Damage" and record on the Y-axis of'the graph on Data Sheet 1, Sheet 2.
5.1.7 Determine the time lapse between core shutdown and the containment area monitor reading and record on the X-axis of the graph on Data Sheet 1, Sheet 2.
5.1.8 Using the monitor reading from Step 5.1.6 and the time lapse, Step 5.1.7, estimate the core damage based on where the values intersect on the graph on Sheet 2 of Data Sheet 1. Record the damage category indicated on Data Sheet 1, Sheet 1.
5.1.9 Obtain the Containment Building atmosphere hydrogen concentration from Procedure 91503-C, " Control Room Instrumentation Output For Assessment Of Core Damage".
5.1.10 Refer to percent hydrogen concentration versus percent Zirconium / Water Reaction Bar Graph on Sheet 2 of Data Sheet 1 to determine the percent of Zircaloy cladding oxidized.
CONTINUED
?c34ag
e-. .
m f PROCEDURE No. REVISloN PAGE NO.
91502-C 0 4 of 46 5.1.11 Record the result on Data Sheet 1, Sheet 1.
CAUTION The bar graph indicating the ,
relationship between the containment hydrogen concentration and the percentage of the Zirconium /
Water Reaction assumes that all of the hydrogen produced in the reaction is released from the reactor coolant to the containment atmosphere. It does not take into account hydrogen depletion due to the operation of the hydrogen recombiners or hydrogen ignitions.
(These factors must be taken into account when attempting to determine the extent of the Zirconium / Water reaction.)
5.1.12 Obtain the CVCS letdown monitor (RE-48000) and containment process radiogas monitor (RE-2565C) from the ERF computer. Record on Data Sheet 1, Sheet 1.
5.1.13 - Complete Preliminary Core Damage Assessment section- of -
Data Sheet 1, Sheet 1. _
5.2- ESTIMATION OF FUEL DAMAGE USING GROSS ANALYSIS DATA 5.2.1 Chemistry personnel obtain post accident samples of reactor coolant system (including reactor sump) and containment atmosphere.
5.2.2 Chemistry personnel also requests isotopic analysis of post accident samples. No decay correction should be applied to sample activities.
NOTE Ensure that sample activities when analyzed are not decay corrected back to the time of sampling as is the routine method of analysis. A decay
! and ingrowth correction factor will be applied to sample activities later in the procedure which will correct I sample activities back to time of reactor shutdown. i l
CONTINUED nus
a f MoCEDURE No. REVISloN PAGE NO.
91502-C 0 5 of 46 5.2.3 When the Isotopic Analysis Report is received from the Chemistry Department, compare nuclides found in the report to those nuclides representing each category of damage in Table 3, " Selected Nuclides For Core Damage Assessment".
NOTE An upper boundary of the extent of fuel damage is made at this point. If indicating nuclides from a given category are not found to be present with specific activities of two orders of magnitude, or more, greater than those normally expected, (see values given in Table 5) it can be interpreted that fuel degradation has_not progressed significantly into that category of damage.
5.2.4 Determine the maximum extent of fuel damage as indicated by the fission products present in samples and record on Data Sheet 1, Sheet 1.
5.3 RADIONUCLIDE SELECTION FOR CORE DAMAGE _ ASSESSMENT _
5.3.1 Select at least two nuclides from each category of damage listed under Table 3, " Selected Nuclides For Core Damage Assessment", for detailed analysis and utilization in final core damage assessment.
NOTE Not all indicating nuclides identified in both the Isotopic Analysis Report and Table 3 are required for this assessment, however, a better overall assessment is made when several nuclides are used. Ensure that nuclides representing each category of damage exhibited in Table 3 and those indicated in Sub-subsections 5.3.2 and 5.3.3 are included.
5.3.2 Include Xenon 133 and at least one other noble gas to allow a noble gas ratio to be determined.
CONTINUED mm
a f PROCEDURE No. REVIStoN PAGE No.
91502-C 0 6 of 46 5.3.3 Include Iodine 131 and at least one other radioiodine to allow a radiciodine ratio to be determined.
NOTES
- a. Consider the length of time that has passed between the reactor shutdown (time of the accident) and the time of sampling and analysis, nuclides with very short half-lives are inappropriate if long periods of time have passed between the reactor shutdown and sample analysis.
- b. Consider the loss of release activity due to a loss of the containment integrity or loss due to condensation prior to the collection of the post accident samples. Reactor
- coolant and containment atmosphere samples drawn after environmental releases occur
'will not accurately represent -- -
total activities released from the core. .
5.3.4 Enter into Column 2, of Data Sheet 2, " Reactor Coolant System Activity", Data Sheet 3, " Containment Atmosphere Activity", and Data Sheet 4, " Additional Activity Released From The Core", the specific activities of the nuclides selected above. Data' Sheet 4 is provided for use under conditions of loss of activity from containment building.
CONTINUED
- C34 4*>
PROCEDURE Mo. REVISION PAGE No.
91502-C 0 7 of 46 NOTES
- a. Data Sheet 4, " Additional Activity Released From Core",
is provided for use in cases where a significant portion of the core fission product activity is contained in sources other than the reactor coolant and containment atmosphere.
- b. Ensure that sample activities are recorded on the worksheets in units of Curies per gram or Curies per cubic centimeter.
If sample results from the Chemistry group are given in units of microcuries, they must be converted to curies before being recorded.
5.4 DETERMINATION OF RELEASE ACTIVITIES 5.4.1 ~
Determine and record the elapsed time from reactor shutdown to the sampl_e count (Isotopic = Analysis) on .-
Data Sheet 4, " Reactor Coolant System Activity", Data Sheet 3, " Containment Atmosphere Activity", and Data Sheet 4, " Additional Activity Released From Core".
5.4.2 For each nuclide selected, determine the ingrowth and decay correction factor using Table 4, " Ingrowth And Decay Correction Factors". Record the factor in Column 3 of Data Sheets 2, 3, and 4.
5.4.3 Multiply the measured specific activities by the ingrowth and decay correction factors and record in Column 4.
5.4.4 Obtain the reactor coolant addition volume and Tavg from Data Sheet 2 " Post Accident RCS Addition Volume Determination", of Procedure 91503-C, " Control Room Instrumentation Output For Assessment Of Core Damage" and record on Data Sheet 2, " Reactor Coolant System Activity".
5.4.5 Convert RCS addition volume, using Data Sheet 2, from gallons to grams mass as follows:
- a. Multiply the total estimated RCS additions volume in line 1 by 3,785 grams per gallon.
- b. Record the grams mass in the appropriate space on line 1.
CONTINUED mm
o e PROCEDURE No. REVISION PAGE No.
91502-C 0 8 of 46 5.4.6 Estimate the RCS mass, using Data Sheet 2, prior to safety injection as follows:
- a. Use 2.45E8 grams if the Reactor Coolant System was from no load temperature and pressure to full load temperature and pressure.
- b. If the Reactor Coolant System temperature was below no load temperature and pressure at the start of the incident, make the following correction:
(1)
From the " Specific Gravity Of Water VS Temperature ' graph, determine the specific gravity of the coolant at Tavg (at the beginning of the incident) and record in the appropriate space on line 2.
(2) Multiply the reactor coolant cool mass (3.32E8 grams), by the specific gravity at Tavg and record on line 2.
5.4.7 Complete the post accident mass calculation, using Data Sheet 2, as follows:
_ Add the pre-safety injection reactor coolant mass a.
from line 1, to the estimated mass from safety injection sources on line 2.
- b. Record the sum on line 3.
5.4.8 Multiply the corrected specific activities, (Column 4) of Data Sheet 2, by the RCS mass and record in Column 5.
5.4.9 On Data Sheet multiply 3, " Containment the corrected specificAtmosphere activities Activity",4)
(Column by the containment atmosphere volume and record in Column 5.
5.4.10 On Data Sheet 4, " Additional Activity Released From Core", if used, multiply the corrected specific activity (Column 4) by the corresponding mass or volume and record in Column 5.
CONTINUED u nus
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O O PROCEDURE No. REVISION PAGE NO.
91502-C 0 9 of 46 5.5 TOTAL RELEASED ACTIVITY OF THE SELECTED RADIONUCLIDES 5.5.1 Record the activities from Column 5 of Data Sheet 2 on the appropriate line of Column 2 of Data Sheet 5,
" Release Activity And Percent Activity Released".
5.5.2 Record the activities from Column 5 of Data Sheet 3 on the appropriate line of Column 3 of Data Sheet 5.
5.5.3 Record the activities from Column 5 of Data Sheet 4 if used, on the appropriate line of Column 4 of Data Sheet 5.
5.5.4 Add Columns 2, 3 and 4 of Data Sheet 5, for each nuclide, and record in Column 5.
5.5.5 Record in Column 6 of Data Sheet 5, the total corrected core inventories from the completed Procedure 91504-C,
" Core Inventory Determinations Using Reactor Power~
History".
5.5.6 Determine the release percentage by dividing the activity of each nuclide on Data Sheet 5, Column 5, by i the total corrected core inventory, Column 6 and multiply by 100.
5.5.7 Record each release percentage in Column 7.
5.6 ESTIMATION ~OF PERCENT FUEL DAMAGE 5.6.1 With the release percentages in Column 7, of Data Sheet 5, enter Enclosure 1, " Instructions For Utilization Of Percent Clad Damage, Pellet Overtemperature And Fuel Melt".
5.6.2 Record the percent damage as indicated by each nuclide on Sheet 3 of Data Sheet 1, " Core Damage Assessment Summary".
i l
l CONTINUED I l
, _ ~ . .
PROCEDURE No. REVISION PAGE NO.
91502-C 0 10 of 46 NOTE Within the limitations of the accuracy associated with this method of assessment, estimates are limited to the following categories:
No Fuel Damage Less Than 50% Cladding Failure Greater Than 50% Cladding Damage Less Than 50% Fuel Overtemperature Greater Than 50% Fuel Overtemperature Less Than 50% Fuel Melt Greater Than 50% Fuel Melt 5.6.3 When attempting to distinguish between NO FUEL DAMAGE and MINOR CLAD FAILURE, perform the following:
- a. Compare the specific activities of the nuclides from the isotopic analysis to those in Table 5,
" Normal Operating Activity". An increase of 100 times or more is considered significant fuel degradation.
~ ~
~' -
-- b. - If the radioiodine activities are -
disproportionally high in comparison to other - :
~
fission product activities, consider that the radioiodine activity may be due to the spiking phenomena and not due to fuel degradation. i
- c. Review Table 6, " Iodine 131 Activity Available For Release Due to Spiking Phenomena", to determine if the iodine activity is elevated due to the spiking phenomena and not fuel degradation.
5.7 ADDITIONAL RADIOLOGICAL INDICATORS 5.7.1 Determine the noble gas ratios by dividing the total corrected release activity of the noble gas nuclides in Column 5 of Data Sheet 5, by the total corrected release activity of Xenon 133.
5.7.2 Compare the noble gas ratios (from Sub-subsection 5.7.1) to those in Table 7, " Noble Gas and Radiolodine Activity Ratios Of Gap and Fuel Pellet".
5.7.3 Record the noble gas ratio under the proper heading on Sheet 1 of Data Sheet 1, " Core Damage Assessment Summary"; GAP activity ratio - record under clad damage category, fuel pellet ratio - record under overtemperature and fuel melt category.
CONTINUED
'c 3 445
PROCEDURE No. REVislON PAGE No.
91502-C 0 11 of 46 5.7.4 Determine the radioiodine activity ratio, divide the total corrected release activity of the radiciodine isotopes in Column 5, of Data Sheet 5, by the total corrected release activity of Iodine 131.
5.7.5 Compare the radiciodine activity ratio to that in Table 7.
5.7.6 Record the radiciodine ratio under the proper heading on Sheet 1 of Data Sheet 1; GAP activity ratio - record as clad damage, fuel pellet ratio - record as overtemperature and fuel melt category.
5.8 FINAL ASSESSMENT 5.8.1 Perform the final core damage assessment by evaluating the data recorded on Data Sheet 1, " Core Damage Assessment Summary", Pages 1, 2, and 3.
NOTE The final assessment is a broad based examination of all data collected. Because of overlapping values of release activities and
~
potential simultaneous conditions -
of clad damage, overtemperature, -
and core melt, Considerable Judgment is required in the final assessment.
6.0 ACCEPTANCE CRITERIA NONE 7.0 REFERENCE 7.1 Westinghouse Owners Group Post Accident Core Damage Assessment Methodology, Revision 2, November, 1984.
7.2 PROCEDURES 7.2.1 91054-C, " Core Inventory' Determinations Using Reactor Power History 7.2.2 91053-C, " Control Room Instrumentation Output For Assessment Of Core Damage" l
l END OF PROCEDURE TEXT l
l vus L
PROCEDURE No. REVISloN PAGE No.
91502-C 0 12 of 46 Sheet 1 of 1 TABLE 1 FUEL DAMAGE VERSUS ELEVATED FUEL ROD TEMPERATURES Fuel Damage Temperature (F)
No Damage <' 13 0 0 Clad Damage 1300 - 2000 Ballooning of zircaloy cladding > 1300 Burst of zircaloy cladding 1300 - 2000 0xidation of cladding and hydrogen > 1600 generation Fuel Ov.ertemperature 2000 - 3450 Fisson product fuel lattice mobility 2000 - 2550 Grain boundry diffusion release of 2450 - 3450 fission products
- ~
"~
Fuel Melt _ _
> 3450
~
Dissolution and liquefaction of uranium > 3450 ~
dioxide in the zircaloy/zirc oxide Melting of remaining uranium dioxide > 5100
Procedure No. Revision Page No. ,
91502-C 0 13 of 46 Sheet 1 of 3 TABLE 2 SUGGESTED SAMPLING LOCATIONS ACCIDENT CONDITION SAMPLINGiLOCATIONS COMMENTS Small Break LOCA Rx Power 4.1% RCS Hot Leg and Containment Atmosphere Alternate RCS samp'le points:
RCS Cold Legs or RCS Pressurizer Rx Power > 1% RCS Hot Leg If RCS is pressurized and i RC pumps are running and Containment Atmosphere Alternate RCS sample points:
RCS Cold Legs or RCS Pressurizer or Residual Heat Removal If RCS on RHR Pump Discharge Large Break LOCA ,
Rx Power < 1% RCS Hot Leg and Containment Atmosphere and Containment Sump Alternate RCS sample points:
RCS Cold Legs or RCS Pressurizer I
simo
i t
Procedure No. Revision Page No. ,
91502-C 0 14 of 46 i Sheet 2 of 3 TABLE 2 SUGGESTED SAMPLING LOCATIONS ACCIDENT CONDITION SAMPLING' LOCATIONS COMMENTS Large Break LOCA '
Rx Power > 1% Containment Atmosphere Provided Containment Sump and Supplying Core Cooling Containment Sump Alternate RCS sample point: Provided that RHR taking Residual Heat Removal Pump from Containment Sump Discharge Steam Line Break RCS Hot Leg and Containmen,t Atmosphere If the containment atmosphere monitor has increased or is increasing Alternate RCS sample points:
- RCS Pressurizer or RCS Cold + Legs Steam Generator Tube Rupture RCS Hot beg and Main Steam and i Containment Atmosphere If the containment atmosphere monitor has increased or is increasing 708610
S' t
Procedure No. Revision Page No. .
91502-C 0 15 of 46 Sheet 3 of 3 TABLE 2 SUGGESTED SAMPLING LOCATIONS g
ACCIDENT CONDITION -SAMPLING LOCATIONS COMMENTS Alternate RCS sample points:
RCS Pressurizer or Alternate Main, Steam sample point:
Main Condenser Hot Well an'd Main Condenser Off Gas naso
I i
e Procedure No. Revision Page No. .
91502-C 0 16 of 46
- Sheet 1 of 1 TABLE 3 SELECTED NUCLIDES FOR CORE DAMAGE ASSESSMENT Core Damage Predominant Gammas, kev Category Nuclide Half-life Yield, (%)
- Clad Failure Kr-85m 4.4 h 150(74), 305(13)
Kr-87 76 m 403(84), 2570(35)
Kr-88 2.8 h 191(35), 850(23), 2400(35)
Xe-131m 11.8 d 164(2)
Xe-133 5.27 d 81(37)
Xe-133m 2.26 d 233(14)
Xe-135 9.14 d 250(91)
I-131 8.05 d 364(82)
I-132 2.26 h 773(89), 955(22), 1400(14) 1-133 20.3 h 530(90)
I-135 6.68 h 1140(37), 1280(34),
1460(12), 1720(19)
Rb-88 17.8 m .
898(13), 1863(21)
Fuel Over-Temperature Cs-134 2y 605(98), 796(99)
Cs-137 30 y 662(85)
Te-129 68.7 m 455(15)
Te-132 77.7 h 230(90)
Fuel Melt Ba-140 12.8 d 537(34) <
La-140 40.22 h 487(40), 815(19), 1596(96)
La-142 92.5 m 650(48), 1910(9), 2410(15) 2550(11)
Pr-144 17.27 m , 695(1.5)
- - Values obtained from Table of Isotopes, Lederer, Hollander, and Perlman, Sixth Edition. .
i a
wo
et I
Procedure No. Revision Page No. .
91502-C 0 17 of 46 Sheet 1 of 2 TABLE 4 INGROWTH AND DECAY CORRECTION FACTORS Kr 85m EXP 0.157t Kr 87- EXP 0.547t Kr 88 EXP 0.247t Xe 131m 1 .
.[-2.62 EXP (-3.54E-3)t] + [3.62 EXP (-2.45E-3)t]
Xe 133 1
[-0.185 EXP-(3.41E-2)t] - [0.10 EXP-(1.28E-2)t + [1.285 EXP-(5.48E-3)t]
Xe 133m 1
[-0.1 EXP -(3.41E-2)t] + [1.1 EXP -(1.28E-2)t} ,
Xe 135 1
[-9.26 EXP-(1.04E-1)t] - [0.033 EXP-(2.66)t] + [10.293 EXP-(7.58E-2)t]
I 131 EXP 0.00359t .
I 132 1 i
[0.103 EXP - (8.91E-3)t] + [0.897tEXP -(0.307)t] ,
I 133 EXP 0.0341t I 135 EXP 0.104t t Cs 134 1 i
70410
r 4
Procedure No. Revision Page No. ,
91502-C 0 18 of 46 Sheet 2 of 2 TABLE 4 '
INGROWTH AND DECAY CORRECTION FACTORS Te 129 1
[1.09 EXP-(0.161)t] + [0.16 EXP-(8.47E-4)t] -
[0.25 EXP-(0.605)t]
Te 132 EXP 0.00892t Ba 140 EXP 0.00225t ,
La 140 1
[1.09 EXP -(2.25E-3)t - [0.09 EXP -(1.72E-2)t] ,
I La 142 1
[-0.14 EXP -(3.78)t] + [ 1.14 EXP :-(0. 44 9) t ]
Pr 144 1
[0.91 EXP -(1.04E-4)t] + [0.09 EXP -(2.4)t] .
- Time, t is the number of hours between shutdown and time of sample count.
L
\
I f ht0
. o PROCEDURE No. REVISION PAGE NO.
41502-C 0 19 of 46 Sheet 1 of 1 TABLE 5 NORMAL OPERATING ACTIVITY
- Reactor Coolant Specific Activity Nuclide (uCi/ gram)
Kr-85m 1.1E-01 Kr-87 6.0E-02 Kr-88 2.0E-01 Xe-131m 1.1E-01 Xe-133 1.8E+01 Xe-135 3.5E-01 1-131 2.7E-01 I-132 1.0E-01
_ I-133- 3.8E-01 1-135 1.9E-01'
- Values obtained from ANS 18.1.
u-
PROCEDURE No. REVISloN PAGE No.
91502-C 0 20 of 46 Sheet 1 of 1 TABLE 6 IODINE 131 ACTIVITY AVAILABLE FOR RELEASE DUE TO SPIKING PHENOMENA Average Reactor Coolant Total I-131 Available I-131 Specific Activity For Release (uCi/ gram) (Curies) 0.5 ( SA ( l.0 3400 0.1 < SA < 0.5 380 0.05 < SA < 0.1 200 0.01 < SA < 0.05 200 0.005 < SA ( 0.01 100 0.001 ( SA < 0.005 100 SA < 0.001 2 Where SA = Normal operating-I-131 specific. activity
_ - (uCi/ gram) in the reactor coolant.
'C344%
PROCEDURE No. REVISION PAGE NO.
91502-C 0 21 of 46 Sheet 1 of 1 TABLE 7 NOBLE GAS AND RADIOIODINE ACTIVITY RATIOS OF GAP AND FUEL PELLET Nuclide Gap Activity Ratio Fuel Pellet Ratio Kr-85m 0.022 0.11 Kr-87 0.022 0.22 Kr-88 0.045 0.29 Xe-131m 0.004 0.004 Xe-133m 0.096 0.14 Xe-135 0.051 0.19 I-132 0.17 1.5 I-133 0.71 2.1
~
I-135 0.39 1.9 Where:
Noble Gas Ratio = Total Activity of Noble Gas Isotope Released Total Activity of Xe-133 Released Radioiodine Ratio = Total Activity of Radiciodine Isotope Released Total Activity of I-131 Released i
t
'03445
{ . . . . . -
~ _. . . .
PROCEDURE No. REVISION PAGE No.
91502-C 0 22 of 46 Sheet 1 of 3 DATA SHEET 1 CORE DAMAGE ASSESSMENT
SUMMARY
- 1. PRELIMINARY CORE DAMAGE ASSESSMENT RVLIS Indication:
No Core Uncovery Uncovery Indicated Length Of Time Uncovered Maximum Core E,xit Thermocouple Reading Fuel Damage Category Based On Thermocouple Temperatures Containment High Range Area Monitor _
Percent Zircaloy/ Water Reaction CVCS Letdown Monitor (RE-48000)
Containment Process Radiogas Monitor (RE-2565C)
= . ,.
DATE: TIME: PERFORMED BY: .
- 2. Record Category Of Fuel Damage Based On The Following Radioisotopic Indications:
Fission Products Present In Samples Fission Product Release Percentages Radiciodine Ratio Noble Gas Ratio DATE: TIME: PERFORMED BY:
7C344%
PROCEDURE MO. REVISIOM PAGE NO.
91502-C 0 23 of 46 Sheet 2 of 3 DATA SHEET 1 CORE DAMAGE ASSESSMENT
SUMMARY
HYDROGEN CONCENTRATION, (V/0) 0.0 2.0 4.0 6.0 8.0 10.0 12.6 l 1 I I l i I I I I I I I I I I I O 10 20 30 40 50 60 70 80 90 100 PERCENT ZIRCONIUM / WATER REACTION, (*/.)
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10 , , , ,,,,,, , , , , -
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10 , , , ,iii., , , ....i. 4 i a si iii, I 10 100 1000 TD2 ATTER SHUTDCW. ROCRS DATE: TIME: PERFORMED BY:
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, o PROCEDURE Wo. REVISlose PAGE No~
91502-C 0 24 of 46 Sheet 3 of 3 DATA SHEET 1 CORE DAMAGE ASSESSMENT
SUMMARY
COLUMN 1 COLUMN 2 COLUMN 3 COLUMN 4 PERCENT CLAD PERCENT PERCENT NUCLIDE DAMAGE OVERTEMPERATURE FUEL MLLT
< 50% 50%> <50% 50% > < 50% 50% >
Kr 85m Kr 87 I
Kr 88 Xe 131m Xe 133 -
Xe 133m Xe 135 I 131 -
~ ~
I I 132 I 133 '
_I 135 Cs 134 1
! Te 129 Te 132
- Ba 140 l
La 140 La 142 Pr 144 Radioiodine Ratio Noble Gas Ratio FINAL ASSESSFENT:
DATE: TIFE : PERFORMED BY:
13J45
=
PROCEDURE NO. REVISION PAGE No.
91502-C 0 25 of 46 Sheet 1 of 2 DATA SHEET 2 REACTOR COOLANT SYSTEM ACTIVITY A. REACTOR MASS DETERMINATION l
- 1. Total
- Gallons X 3,785 Gram / Gallon = Grams
- 2. Enter 2.45E8 Grams if RCS was no load temperature and pressure to full load temperature and pressure. If less than no load temperature and pressure, use the following correction:
3.32E8 Grams X (Specific Gravity of = + Grams '
RCS At Tavg*) l t .. 3. Total Post Accident Reactor Coolant System = Grams Mass
~ '
SPECIFIC GRAVITY -- -
0.75 1.0 I i I i i I 6 30 400 100 WATER TEMPERATURE, (*F)
- From 91503-C - Data Sheet 2
?C344*
PROCEDURE NO. REVISION PAGE NO.
91502-C 0 26 of 46 Sheet 2 of 2 DATA SHEET 2 B. NUCLIDE CALCULATION ELAPSED TIME, HOURS, SHUTDOWN TO SAMPLE COUNT COL 1 COL 2 COL 3 COL 4 COL 5 MEASURED INGROWTH / DECAY INGROWTH / DECAY NUCLIDE SPECIFIC CORRECTION CORRECTED RCS ACTIVITY, FACTOR SPECIFIC ACTIVITY ACTIVITY, CI/ GRAM CI/ GRAM CI Kr 85m Kr 87 Kr 88 Xe 131m Xe 133 Xe 135
~
~
I 131 I 132 1 133 I 135 Cs 134 Te 129 Te 132 Ba 140 La 140 La 142 Pr 144 l
l DATE: TIME: PERFORMED BY:
- mus ;
PROCEDURE NO. REVISIOM PAGE MO.
91502-C 0 27 of 46 Sheet 1 of 1 DATA SHEET 3 CONTAINMENT ATMOSPHERE ACTIVITY ELAPSED TIME, HOURS, CONTAINMENT VOL, (7.9 E10 CC)
SHUTDOWN TO SAMPLE COUNT COL 1 COL 2 COL 3 COL 4 COL 5 MEASURED INGROWTH / DECAY INGROWTH / DECAY CONTAINMEt;T NUCLIDE SPECIFIC CORRECTION CORRECTED ACTIVITY, ACTIVITY, FACTOR SPECIFIC ACTIVITY CI CI/CC CI/CC Kr 85 Kr 85m Kr 87 Kr 88 Xe 131m
_ Xe 133 -= -
~- -
Xe 135 I 131 1 132 1 133 I 135 Cs 134 Te 129 Te 132 Ba 140 La 140 I La 142 Pr 144 DATE: TIME: PERFORMED BY:
~nm
PROCEDURE NO. REVISION PAGE NO.
91502-C 0 28 of 46 Sheet 1 of 1 DATA SHEET 4 ADDITIONAL ACTIVITY RELEASED FROM CORE ELAPSED TIME, HOURS, MASS OR VOLUME,GM/CC SHUTDOWN TO SAMPLE COUNT COL 1 COL 2 COL 3 COL 4 COL 5 MEASURED INGROWTH / DECAY INGROWTH / DECAY ACTIVITY NUCLIDE SPECIFIC CORRECTION CORRECTED RELEASED, ACTIVITY, FACTOR SPECIFIC ACTIVITY CI CI/G(CC) CI/G(CC)
Kr 85 Kr 85m Kr 87 Kr 88 Xe 131m
=-Xe 133 - - -
Xe 135 I 131 1 132 I 133 I 135 Cs 134 Te 129 Te 132 Ba 140 La 140 La 142 Pr 144 DATE: TIME: PERFORMED BY:
703445
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PROCEDURE No. REVISloN PAGE NO.
91502-C 0 31 of 46 Sheet 1 of 2 ENCLOSURE 1 INSTRUCTIONS FOR UTILIZATION OF PERCENT CLAD DAMAGE, PELLET OVERTEMPERATURE, AND FUEL MELT 1.0 To determine percent clad damage select correct set of graphs as follows:
1.1 If the core burnup is less than 5,000 MWD /MTU, use bottom line of appropriate graph.
1.2 If the core burnup is between 5,000 and 25,000 MWD /MTU, use middle line of appropriate graph.
1.3 If the core burnup is greater than 25,000 MWD /MTU, use top line of appropriate graph.
1.4 If iodine spiking is indicated use appropriate graph.
1.5 Obtain release percentage values from column 7 of Data Sheet 5, " Release Activity and Percentage Activity.-
~
Releised". -
1.6 Enter each graph with the corresponding release percentage value to obtain the percent clad damage estimate. Record the percentage clad damage values on Data Sheet 1, " Core Damage Assessment".
2.0 To determine percent fuel overtemperature select the correct set of bar graphs ss follows:
i 2.1 If the core burnup"is less than 5,000 MWD /MTU, use " Low Burn-up Bar Graphs .
2.2 If the core burnup is between 5,000 and 25,000 MWD /MTU, use " Average Burn-up Bar Graphs".
2.3 If the core burnup is greater than 25,000 MWD /MTU use "High Burn-up Bar Graphs l
l l
i D44$
PROCEDURE NO. REVISloN PAGE No.
91502-C 0 32 of 46 Sheet 2 of 2 ENCLOSURE 1 (Cont'd.)
2.4 Obtain percent inventory released values from column 7 of Data Sheet 5, " Release Activity and Percentage Activity Released".
2.5 Enter each bar graph with the corresponding release percentage value to obtain the percentage of fuel overtemperature estimate. Record the percentage fuel overtemperature value on Data Sheet 1, " Core Damage Assessment".
3.0 To determine the percent fuel melt select correct set of bar graphs as follows:
3.1 If the core burnup is less than 5,000 MWD /MTU use " Low Burn-up Graphs".
3.2 If the core burnup is between 5,000 and 25,000 MWD /MTU, use " Average Burn-up Bar Graphs".
3.3 Obtain percent inventory released values from column 7 of Data Sheet 5, " Release Activity and Percentage Activity Released".
~
3,. 4 -- Enter each bar graph with the corresponding release percentage value to obtain the percent fuel melt estimate. Record the percentage fuel melt values on Data Sheet 1, " Core Damage Assessment".
703445
- g. .
---P p;,0CEDURE NO. REVISION PAGE NO.
91502-C 0 33 Of 46 s
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RELATIONSHIP 0F % CLAD DAFAGE WITH % CORE INVENTORY RELEASED OF KR-87 uns
PROCEDURE NO. REVISION PAGE NO 91502-C 0 34 Of 46 1- ! ; ;:m ; ; i f :+3 '
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RELATION OF % CLAD DAMAGE WITH 1 CORE INVENTORY RELEASED OF XE-131M
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PROCEDURE NO. REVISION PAGE MO.
91502-C 0 35 of 46 A
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RELATION OF i CLA0 DAMAGE WITH CORE INVENTORY RELEASED OF RE-133 4
'03445
PROCEDURE NO, REVISION PAGE NO.
91502-C 0 36 of 46 1 i i i : li ! ; ; i: :: ; ; ; ;;..,
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r RELATIONSHIP 0F ', CLAD DAFAGE WITH 's CORE INVENTORY RELEASED OF I.131
.i
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PROCEDURE NO. REVISION PAGE NO.
91502-C 0 37 of 46 r
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RELATIONSHIP 0F % CLAD DAMAGE WITH % CORE INVENTORY RELEASED OF- I-132 m..s
PROCEDURE NO. REVISION PAGE NO.
91502-C 0 38 Of 46 10'
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1 1
l RELATIONSHIP 0F i CLAD DAMAGE WITH % CORE INVENTORY RELEASED OF 1-133 1
- 03J45
PROCEDURE NO. REVISION PAGE NO.
91502-c 0 39 of 46 f
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PROCEDURE NO. REVISION PAGE NO 91502-C 0 40 of 46 10' '
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RELATIONSHIP OF : CLAD DAMAGE WITH % CORE INVENTORY RELEASED OF I-131 WITH SPIKING w.s
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o J PROCEDURE NO. REVISION PAGE NO.
91502-C 0 44 of 46 PERCENT CORE INVENTORY RELEASED VERSUS PERCENT FUEL MELT FOR LOW BURN-UP CASE Core Inventory Released, (%) For Kr, Xe, I, Cs and Te W-S 8 8 8, 8 O
~ m . S. O- ~ O O. C. ~0 n
l S 8 88 O O C0
- ~ n 8 S. O
. - ~ O n O. . .
Fuel Melt, (%)
Core Inventory Released, (%) For Ba and Sr 2 *
- W W S, S. SS.- ~_. n _, .
O
~
l I I l~ll I I I I Illi I l l 8 .S SO O. C0 S
~ S n S, O O O,
- .- ~ n .-
. Fuel Melt, (%)
- Core Inventory Released, (%) For Pr 7 7 7 7*7 7 7 7 77 W W W W WWW W W W WW l l I I l~ l I ll 1 11 1I S S S S, S SO O O O 00
- ~ n . .- ~ n O. . -
Fuel Melt, (%)
J
' ' ~ '
,J-PROCEDURE NO. REWSION PAGE Ro.
91502-C 0 45 of 46 t
l PERCENT CORE INVENTORY RELEASED VERSUS
! PERCENT FUEL MELT FOR AVERAGE BURN-UP CASE Core Inventory Released, (7.) For Kr, Xe, I, Cs and Te o o e o e o- - -
U E M W W WW W W W $$
I I I I I ll 1 I I I I Ill I o o o e o e- - - -
U U M W W WW W WW $ $U ~
Fuel Melt, (7.)
l Core Inventory Released, (7.) For Ba and Sr O
m O, O08 m --
.m 8 8 m 8, 8 8. = c
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Fuel Melt, (7.)
Core Inventory Released, (7.) For Pr 77 7 7T T To M M W W WM M MW W MU
'l Ill i I I I 'I til l
e 0 0 0 =0 m a, e, em C m -
S. 8. ~ m e Fuel Melt, (7.)
m.s
c-.
PROCEDURE NO. REVISION PAGE NO 91502-C 0 46 of 46 PERCENT CORE INVENTORY RELEASED VERSUS PERCENT FUEL MELT FOR HIGH BURN-UP CASE Core Inventory Released, (7.) For Kr, Xe , I, Cs and Te S
8
~ W, 8 88C
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Core Inventory Released, (7.) For Ba and Sr O
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l l
l
)34Sb
7_.
- A1prov11 Vogtle Electric Generating Plant 91503-C NUCLEAR OPERATIONS RJvision No.
/ [ o cate
""" '*" Georgia Power e.se no.
O 6 a es- 1 of 4 CONTROL ROOM INSTRUMENTATION OUTPUT FOR ASSESSMENT OF CORE DAMAG 1.0 PURPOSE This procedure provides instruction for collecting and recording information obtained from Control Room instrumentation needed in sessi ' "thhlextent of core y)$U damage. U%
f69 t 2.0 PRECAUTIONS AND' M MITA .,
NONE 3.0 PREREQUISITES An emergency condition has been declared and core damage is suspected and Emergency Director has directed O the implementation of this procedure.
d 4.0 RESPONSIBILITIES 4.1 OPERATIONS DEPARTMENT 4.1.1 The Operations Department shall execute this procedure and transmit the recorded results to the Health Physics personnel coordinating core damage assessment activities.
5.0 MAIN BODY 5.1 RVLIS READINGS AND RECORDING 5.1.1 Review the Reactor Vessel Level Instrumentation System, (RVLIS) indications uncovered at any timeto determine during the iftransient.
the core was If it is apparent that it was uncovered, estimate the length of time that it was uncovered and record on DATA SHEET 1,
" Control Room Instrumentation Data Record For Core Damage Assessment".
O CONTINUED 703'01 A mim
r e
PAGE No PROCEDURE No. REVISION 0 2 of 4 91503-C 0 5.2 CORE EXIT THERMOCOUPLE TEMPERATURES 5.2.2 Record on DATA SHEET 1, " Control Room Instrumentation Data Record For Core Damage Assessment", all
~
temperatures exceeding 1300 degrees Fahrenheit along with the corresponding thermocouple identification numbers. A core map is provided for additional use in display of the location of thermocouples indicating high temperatures.
5.3 RADIATION MONITOR READINGS AND RECORDS 5.3.1 Determine the maximum output value from the containment high range monitor and containment atmosphere hydrogen monitor.
5.3.2 Record the values obtained in the spaces provided on DATA SHEET 1.
5.4 Determine volume of all Reactor Coolant System (RCS) additions made during the accident and prior to the collection of RCS core damage assessment samples by the following:
- a. From Control Room Tank Level indications prior to
(~)'
and following safety injection, estimate the volume of each addition, and convert to gallons.
- b. Record the estimated addition for each source on the appropriate line in DATA SHEET 2, " Post Accident RCS Volume Determination".
- c. Record initial T,yg at commencement of transient.
5.5 Transmit the completed DATA SHEETS to the Health Physics personnel coordinating core damage assessment activities.
6.0 ACCEPTANCE CRITERIA NONE
7.0 REFERENCES
7.1 Westinghouse Owners Group Post Accident Core Damage Assessment Methodology, Revision 2, November, 1984.
7.2 PROCEDURES 91502-C, " Core Damage Assessment" 7.2.1
{) 7.2.2 91504-C, " Core Inventory' Determinations Using Reactor Power History END OF PROCEDURE TEXT
PAGE NO R E'MSION PROCEDUREf40 3 of 4 0
91503-C sneet 1 or i O
V DATA SHEET 1 CONTROL ROOM INSTRUMENTATION DATA RECORD FOR CORE DAMAGE ASSESSMENT Time: Performed by Date:
Length Of Time Core Uncovered Maximum Containment Bldg.
Containment Building High Range Monitor Output Atmosphere Hydrogen (%)
1 2 .
3 4
\
(U 5 6
7 .
8 .
9 .
10 .
11 .
12 .
13 14 15 D C B A L K J H G F E R P N M 1300 Degrees F Thermocouple Temperatures
.- J
PAGE NO REVISION PROCEDUREfio 4 of 4 0
91503-C Sheet 1 or 1 DATA SHEET 2 POST ACCIDENT RCS ADDITION VOLUME DETERMINATION Time: Performed by Date:
- 1. Source Of Tank Level. (%) Addition Pre-Accident Post Accident (From Daily Actual Delta Conversion Gallons Level Factor Added Logs) (I) % Level (II) Z (II-I) %
RWST Boric Acid Tk Accumulator Accumulator Accumulator Accumulator TOTAL GALLONS ADDED
- 2. Record T,yg at the commencement of transient O
Procedur3 No.
Approv;p Vogtle Electric Generating Plant 91504-C NUCLEAR OPERATIONS Revision No.
0 at.
-C (o 6
/ /- Unit COMMON Georgia Power e.s. No.
1 of 21 h$$ : ':. p
)
i CORE INVENTORY DETERMINATIONS USING REACTOR POWER HISTORY 1.0 PURPOSE This procedure provides instructions and DATA SHEETS necessary for manual determination of. fission product core inventories. These, estimate's? based'on'g,ldamage Reactor Power History, are ;usiid tih%p'ost!'acci'degt' assessment and represet, fehN accident.
, time of reactorshutdownpr@ git @
re damage 2.0 PRECAUTIONS AND LIMITATIONS 2.1 The reactor power fluctuations of no more than plus or minus ten percent of the recorded power level values, are acceptable within any power level period.
/ 2.2 The length of Reactor Power History required for determining core inventories varies with each nuclide used in the assessment. A reactor power history equal to at least four (4) times the half-life of the nuclide is required and is specified below Column 2 of each Reactor Power History DATA SHEET.
i
- 3.0 PREREQUISITES .
An emergency condition has been declared and core damage is suspected and the Emergency Director has directed the implementation of this procedure.
4.0 RESPONSIBILITIES 4.1 It is the responsibility of the Engineering Department to execute this arocedure and to transmit the recorded results to Health Physics personnel coordinating the assessment actions.
CONTINUED
e e
PAGE No.
REVISION P<,oCEDURE No.
0 2 of 21 91504-C 0 5.0 MAIN BODY 5.1 INSTRUCTIONS full 5.1.1 Obtain Reactor Power History, in unirs of percent rated power from Emergency Response Facility (ERF) computer. If the information is unavailable from ERF computer, obtain power level values from the control room.
5.1.2 Enter reactor power level as percent of rated full power in Column 1 of DATA SHEETS 1 through 18.
5.1.3 For each percent power level valua entered in Column 1 of DATA SHEETS 1 through 11 and 13 through 18, enterThe the duration at this power, in hours, in Column 2.
power level and duration values should be for time period equal to or slightly greater than the minimum time indicated below Column 2 to be used for calculation purposes.
5.1.4 DATA SHEET 12 is used to determine the power correction factor for Cs-134. To use DATA SHEET 12, the average power during the entire operating period is required.
(Due to the production characteristics of Cs-134, a f]).
' different method must be used in determining the power correction factor.)
5.1.5 Enter into Column 3, of each DATA SHEET, the time between the end of each power period recorded in Column 1 and the time at which reactor shutdown occurred, in hours.
5.1.6 Using the bar graph provided for each nuclide, determine the Factor A associated with each time value in Column 3. Record each Factor A in Column 4.
5.1.7 Using bar graph provided for each nuclide, determine the Factor B associated with each time value, recorded in Column 2. Record each Factor B in Column 5.
5.1.8 Multiply each respective power level (Column 1) with its corresponding Factor A (Column 4) and Factor B (Column 5), to obtain the product of percent power.
Enter the product in Column 6. I f
I 5.1.9 Add the products in Column 6 to determine the power correction factor and enter the value below Column 6.
3 (d
CONTINUED 703445
PAGE No.
REVIStoH PROCEDURE NO.
0 3 of 21 91504-C O 5.1.10 Calculate the core inventory for each nuclide by multiplying the power correction factor from Sub-subsection 5.1.8 by the respective nuclide factor given on the DATA SHEET.
5.1.11 Enter the product obtained on the line labeled " Total Corrected Activity 5.1.12 Transmit the completed DATA SHEETS to the Health Physics personnel responsible for coordinating the core damage assessment activities.
6.0 REFERENCES
6.1 Westinghouse Owner's Group Post Accident Core Damage Assessment Methodology, Revision 2, November, 1984.
l 6.2 PROCEDURES 6.2.1 91502-C, " Core Damage Assessment" 6.2.2 91503-C, " Control Room Instrumentation Output For Assessment of Core Damage" O END OF PROCEDURE TEXT i
9 O
~
REVISION PAGE NO.
PROCEDURE NO. 4 of 21 91504-C 0 O DATA SHEET 1 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: Kr 85m Column 2 Column 3 Column 4 Column 5 Column 6 Column 1 Factor Product Average Duration Shutdown Factor At Power To End Of A B %P x A x B Power Level Power (I P) (Hours) (Hours)
O
- Power Duration = Correction Factor PgxA1xB1=
i
- Must be at least 17.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Factor A 2 1.5 1 0.7 0.5 0, 3 10 9 8 7 6 5 4 3 I I I I I I I 1 i g g g g g g g
' 3 1 3 I I I I I I I I e i i i i i ' '-
Time From Column 3 (Hours) 1 13 14 15 16 17 18 19 20 21 22' ime F om Co umn 2 (Houd) 10 11 '12 TTI I l l l I I l 1 l I
l l l I l l l l l
l j
l 3 I I I I I I - I 3 0 1 2 3 4 5 6 7 y a g,3 g.3 9.5 p.7 Factor B Core Inventory = 2.10E3 X = Curies PCF, col 6 sum Total Corrected Core Activity TIME: PERFORMED BY:
DATE:
'o 2" 5
PAGE NO.
REVISloN PROCEDURE NO.
0 5 of 21 91504-C O DATA SHEET 2 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: Kr 87 Column 4 Column 5 Column 6 Column 1 Column 2 Column 3 Product Shutdown Factor Factor Average Duration A B %P x A x B Power At Power To End Of Level Power (I P) (Hours) (Hours)
O
- Power Duration = Correction Factor PgxAtxBt=
i
- Must be at least 5.0 hours.
Factor A 2 1.5 1 0.7 0.5 0< 3 6 5 4 3 10 9 8 7 l 1 i j 1 I l I I I I f 1 I I I 1 l l 1 1 I I l l l 3 l Time From Column 3 (Hours) 1 3 '4 5 '6 fimeFromkolumn2fHours) I i i I I I I 1 I i 1 1 i i l I i i 1 I I I I i i I I I 9.5 1.7 8 9.0 9.3 0 1 2 3 4 5 6 7 Factor B
= Curies Core Inventory = 3.83E3 X Total Corrected PCF, col 6 sum Core Activity TIME: PERFORMED BY:
DATE:
mco
REVISloN PAGE NO.
PROCEDURE No.
91504-C 0 6 of 21 O DATA SHEET 3 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: Kr 88 Column 2 Column 3 Column 4 Column 5 Column 6 Column 1 Factor Product Average Duration Shutdown Factor At Power To End Of A B %P x A x B Power Level Power (I P) (Hours) (Hours)
O
- Power Duration = Correction Factor P gxA fxB i=
i
- Must be at least 11.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Factor A 2 1.5 1 0.7 0.5 0, 3 10 9 8 7 6 5 4 3 I I I I I I I i ! I 1 ! 1 I I I l g I I I i l [ [ l l l Time From Column 3 (Hours) l 6 7 8 9 10 11 12 13 14 fimeFromColumn2(Houhs) l l l , l 1 1 1 1 1 t 1 I I i i i i i I s i i i
' I 1 1 I 8 9.0 9.3 9.5 ).7
(' I 2 3 4 5 6 7 Factor B O Core Inventory = 5.45E3 X PCF, col 6 sum
=
Total Corrected Curies Core Activity TIME: PERFORMED BY:
DATE:
703445
PAGE NO, PROCEDURE No. REVISION 91504-C 0 7 of 21 O DATA SHEET 4 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: Xe 131m Column 2 Column 3 Column 4 Column 5 Column 6 Column 1 Factor Product Average Duration Shutdown Factor Power At Power To End Of A B %P x A x B Level Power (I P) (Hours) (Hours)
O
- Power Duration = Correction Factor P g xA gxB t=
i
- Must be at least 1132 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.30726e-4 months <br /> Factor A 2 1.5 1 0.7 0.5 0, 3 10 9 8 7 6 5 4 3 1 I i 1 i i i I I i l i I I I I I I I I I I I I l I I I I Time From Column 3 (Hours)I 1
8 0 200 4 800 1000 1200 1400 Time From Column 2 ll00 Hours) 600 I I I I I I I I I I I I l I I I I I I I l i I I I I I I i i 8 9.0 9.3 9.5 p.7 0 1 2 3 4 5 6 7 Factor B O Core Inventory = 6.03El X = Curies PCF, col 6 sum Total Corrected Core Activity DATE: TIME: PERFORMED BY:
mas
REVISION PAGE NO.
PROCEoVRE NO.
0 8 of 21 91504-C DATA SHEET 5 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: Xe 133 Column 2 Column 3 Column 4 Column 5 Column 6 Column 1 Factor Product Average Duration Shutdown Factor At Power To End Of A B %P x A x B Power Level Power (I P) (Hours) (Hours) _
i A
L)
- Power Duration = Correction Factor PfxAfxB 1=
i
- Must be at least 506 hours0.00586 days <br />0.141 hours <br />8.366402e-4 weeks <br />1.92533e-4 months <br /> Factor A 1.5 1 0.7 0.5 0,3 6 5 4 3 2 10 9 4 7 I i l i i i l i I I I I i i I I I I I l 1 Time From Column 3 (IHours) '
I I
I 200 '300 '400 '500 600 3100' p'me From Column 2 (Hours)
I I I i 1
i I I i i I i i I i i I I i 1 6 7 8 9.0 9.3 9.5 p./
() 1 2 3 4 5 Factor B
= Curies Core Inventory = 1.91E4 X Total Corrected PCF, col 6 sum Core Activity TIME: PERFORMED BY:
DATE:
7C3445
REVISION PAGE NO.
PROCEDURE No.
91504-C 0 9 of 21 O DATA SHEET 6 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: Xe 133m Column 2 Column 3 Column 4 Column 5 Column 6 Column 1 Product Average Duration Shutdown Factor Factor Power At Power To End Of A B %P x A x B Level Power (I P) (Hours) (Hours)
O
- Power Duration = Correction Factor PtxAtxB t=
i
- Must be at least 217 hours0.00251 days <br />0.0603 hours <br />3.587963e-4 weeks <br />8.25685e-5 months <br /> Factor A 2 1.5 1 0.7 0.5 0, 3 10 9 8 7 6 f 4 3 I i I I I I I I I I I I I I 1 1 I I I I I I I I I I I I i l 1 1 I I I i 1 l l l Time From Column 3 (Hours)I I I I I I I I I I I I I I ' I g I I 50 10d 15d 20d 250 Time From Column 2 (Hours)
I I I I I I I I I I I I I I i 1 I I I l l l l l l l I I I I I I I I I I I I I I I I 9.0 9.3 9.5 p.7 O1 2 3 4 5 6 7 8 Factor B O Core Inventory = 2.68E3 X = Curies PCF, col 6 sum Total Corrected Core Activity DATE: TIME: PERFORMED BY:
mus
w-PAGE No.
REVISloN PROCEDURE NO.
91504-C 0 10 of 21 G
V DATA SHEET 7 REACTOR. POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: Xe 135 Column 2 Column 3 Column 4 Column 5 Column 6 Column 1 Factor Product Average Duration Shutdown Factor Power At Power To End Of A B %PxAxB Level Power (I P) (Hours) (Hours; O
- Power Duration = Correction Factor PfxAgxB 1=
- Must be at least 36.6 hours Factor A 2 1.5 1 0.7 0.5 0,3 le 9 8 7 6 5 4 3 I I i i I il l 1 1 I i f I 1 1 l l l I I I l I I ill II III Time From Column 3 (Hours) 1 I 1 I
10 20 30 40 0
Time From Column 2 (Hours)
II iti I1 III I I I i #
I i i I I I I I I i i i I I I i 9.3 9.5 p.7 O 1 2 3 4 5 6 7 8 9.0 Factor B b) v Curies Core Inventory = 3.54E3 X =
Total Corrected PCF, col 6 sum Core Activity DATE: TIME: PERFORMED BY:
7C3445
REVISION PAGE NO.
PROCEDURE NO.
91504-C 0 11 of 21 O
V DATA SHEET 8 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: I 131 Column 1 Column 2 Column 3 Column 4 Column 5 Column 6 Average Duration Shutdown Factor Factor Product Power At Power To End Of A B %P x A x B Level Power (I P) (Hours) (Hours)
O
- Power Duration = Correction Factor P xA 1 xB 1=
t i
- Must be at least 773 hours Factor A 2 1.5 1 0.7 0.5 u,3 10 9 8 7 6 5 4 3 I I I I i i 1 i i I I I I I I I I I I I i i i Time From Column 3 Hours)i 00 700 800 900 fimeFro Column 2(hurs) 1 I l i I I i 1 I I I I I I I I I I I I I I I I I O 1 2 3 4 5 6 7 8 9.0 - 9.3 9.5 .l . 7 I
Factor B A
V Curies Core Inventory = 9.38E3 X =
PCF, col 6 sum Total Corrected Core Activity DATE: TIME: PERFORMED BY:
703445
PAGE NO.
REVIStoN PRGCEDURE No.
0 12 of 21 91504-C O DATA SHEET 9 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: I 132 Column 3 Column 4 Column 5 Column 6 Column 1 Column 2 Product Duration Shutdown Factor Factor Average B %P x A x B At Power To End Of A Power Level Power (I P) (Hours) (Hours)
- Power Duration = _ . ,
Correction Factor PtxAtxBt=
i
- Must be at least 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> Factor A 0.7 0.5 0, 3 6 5 4 3 2 1.5 1 10 9 8 7 i i i i i i i i i i i i i i e i I l l 1 l l 1 i I i I Time From Column 3 (Hours) I i l
i I t i I I '
I I I3 '4 '5 I
6 '7 '8 '9 10 'l1 0 I 2 Time From Column 2 (Hours)
I I I I I l l 1 l I I I I I I I I i I i 1 I I I I I 1
8 94 9.3 9.5 J.7 0 1 2 3 4 5 6 7 Factor B f (m.
= Curies 1 Core Inventory = 1.34E4 X Total Corrected I
PCF, col 6 sum Core Activity TIME: PERFORMED BY:
DATE:
733445
REVISloN PAGE NO.
PROCEDURE NO.
91504-C 0 13 of 21 O DATA SHEET 10 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: I 133 Column 2 Column 3 Column 4 Column 5 Column 6 Column 1 Factor Product Average Duration Shutdown Factor Power At Power To End Of A B %P x A x B Level Power (I P) (Hours) (Hours)
O J
- Power Duration = Correction Factor P xA fxB f=
f
- Must be at least 81_ hours Factor A 1.5 1 0.7 0.5 0,3 5 4 3 2 10 9 8 7 6 I I l I I I I I I I I I i 1 1 I i l i L i i I I I l
Time From Co'.umn 3 (Hours) I I I i I I 3
3 3 3 50 '60 70 '80 90 100 O 10 20 0 Time From Column 2 Hou rs )40 I I I i I I l I 1 1 I I I i I I I I I I I I I i ( I
).7 7 8 9.0 9.3 9.5
( 1 2 3 4 5 6 Factor B Curies Core Inventory = 1.91E4 X =
PCF, col 6 sum Total Corrected Core Activity DATE: TIME: PERFORMED BY:
703445
REVISION PAGE No.
PROCEDURE NO.
91504-C 0 14 of 21 DATA SHEET 11 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: I 135 Column 2 Column 3 Column 4 Column 5 Column 6 Column 1 Factor Product Average Duration Shutdown Factor Power At Power To End Of A B %P x A x B Level Power (I P) (Hours) (Hours)
O
- Power Duration = Correction Factor P(xAfxBg=
i
- Must be at least E hours Factor A 4 3 2 1.5 1 0.7 0.5 04 3 10 9 8 7 6 5 I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I i i i I I I I I I I II I I Time From Column 3 (Hours) I
'25I I 'I I I I I I I I I I I 10' ' I I 1
' 30
! O I I ' I 5' ' '
15 20 Time From Column 2 (Rours)
I I i I i i I I I I II I I I I I I I I I I I I I II I I I I I i i I i i i l l i I I I I i i i ).7 8 90 9.3 9.5 O 1 2 3 4 5 6 7 Factor B l
O Core Inventory = 1.72E4 X =
g Curtee Core Activity l
DATE: TIME: PERFORMED BY:
i mus
REVISION PAGE NO.
PROCEDURE NO.
91504-C 0 15 of 21 O DATA SHEET 12 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: Cs 134 t.e 90%
Power 8.9 Es -
- 75%
- Power 8.7 8.. . ,
e.s 50%
Power L4 r
/ .
O u -
8.3 ,
8.1 "8 as a ses ase taan e
1 i
l CYCLE OPERATION (CALENDAR DAYS) l NOTE Must be at lease 34 days of cycle operation. 1 l
Curies Core Inventory = 2.81E7 X ( ) =
PCF Total Corrected Core Activity O
DATE: TIME: PERFORMED BY:
I T3344$
l .
REVISION PAGE NO.
PROCEDURE No.
91504-C 0 16 of 21 O DATA SHEET 13 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: Te 129 Column 3 Column 4 Column 5 Column 6 Column 1 Column 2 Product Duration Shutdown Factor Factor Average B %PxAxB Power At Power To End Of A Level Power (I F) (Hours) (Hours)
O N
- Power Duration = Correction Factor PgxAgxB g=
i
- Must be at least 5 hours Factor A 0.7 0.5 0, 3 6 5 4 3 2 1.5 1 10 9 8 7 I I I I I I I I I I I I I 1 1 1 I I I I I I 1 l 1 I Time From Column 3 (Hours) 1 I
I 2 3 4 5 O 1 Time From Column 2 (Hours)
I I I I I I I I i I I I I I I I I I I I I I I I I I ( 9.5 ).7 8 9.0 9.3 0 1 2 3 4 5 6 7 Factor B O Core Inventory = 3.16E3 X =. Curies 4
PCF, col 6 sum Total Corrected
- Core Activity l
TIME: PERFORMED BY:
DATE:
ma5 . . . _ _ _ _ _ _ . _ ._ _ _ _ _ _ _ . . _ - _ - __, _ _ . _
PAGE NO.
REVISION PROCEDURE NO.
91504-C 0 17 of 21 l
\
O DATA SHEET 14 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION i '
NUCLIDE: Te 132 Column 3 Column 4 Column 5 Column 6 Column 1 Column 2 Product Duration Shutdown Factor Factor Average B %P x A x B s
- Power At Power To End Of A i Level Power (I F) (Hours) (Hours) i i,
O
- Power Duration = Correction Factor P g xA gxB 1=
i
- Must be at least 310 hours Factor A 0.7 0.5 0,3 6 5 4 3 2 1.5 1 10 9 8 7 i i i i 1 I i i i i i i i i 1 1 I I I I I I I I I I I I I I Time From Column 3 (Hours) i I I I 300 I ' I I I I 10 200 D
Time From Column 2 Hours) i 1 I I i II I I II I I I I I I I I I I I I I I i l l 9.3 9.5 1.7 6 7 8 9.0 O 1 2 3 4 5 Factor B O Core Inventory = 1.34E3 X = Curies PCF, col 6 sum Total Corrected i Core Activity TIME: PERFORMED BY:
DATE:
-- raus.- . __ _ -
REVISloN PAGE NO.
PROCEDURE No.
91504-C 0 18 of 21 O DATA SHEET 15 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: Ba 140 Column 2 Column 3 Column 4 Column 5 Column 6 Column 1 Factor Product Average Duration Shutdown Factor At Power To End Of A B %P x A x B Power Level Power (I P) (Hours) (Hours)
O -
- Power Duration = Correction Factor PfxAfxBy=
i
- Must be at least 1230 hours Facco,r A 3 2 1.5 1 0.7 0.5 04 3 10 9 8 7 6 5 4 I I I I I I I i 1 1 1 1 I I I 1
I I I I I I I l I I I I I I I Time From Column 3 (Hours) 40 00 800 1000 1200 1400 0 200 Time From Column 2 Hours
! I I I i i I i 1 1 I i I I I I I I I I I I I I I I I I I I I 9.0 9.3 9.5 3.7 5 6 7 8 (i 1 2 3 4 Factor B O Core Inventory = 1.63E4 X = Curies PCF, col 6 sum Total Corrected Core Activity DATE: TIME: PERFORMED BY:
703445 , ,-
PAGE No.
REVISION PROCEDURE NO.
0 19 of 21 91504-C (a~'s DATA SHEET 16 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: La 140 Column 3 Column 4 Column 5 Column 6 Column 1 Column 2 Factor Product Average Duration Shutdown Factor To End Of A B %P xAxB Power At Power Level Power (I P) (Hours) (Hours) m
.)
- Power Duration = Correction Factor PgxAgxBy=
i
- Must be at least 160 ,
hours Factor A 0.7 0.5 0, 3 5 4 3 2 1.5 1 10 9 8 7 6 i 1 1 1 1 1 1 i i i i i 1 i i i i l I l l l l l l l I I I I I I i i i Time From Column 3 (Hours) -
140' 160 180 200
'40 100 '120 20 f' ime From Column 2 (Yours)80 I I I I I I I I I I I i l I I I I I I I I I I I I I I I I I I I I I I I J.7 8 9.0 9.3 9.5 4 6 7 O 1 2 3 5 Factor B
= Curies Core Inventory = 1.72E4 X Total Corrected PCF, col 6 sum Core Activity TIME: PERFORMED BY:
DATE:
on _ ___. ___ __ _. _ - _ - -
REVISION PAGE NO.
PROCEDURE NO.
0 20 of 21 91504-C O DATA SHEET 17 REACTOR POWER HISTORY AND CORE INVENTORY DETERMINATION NUCLIDE: La 142 Column 2 Column 3 Column 4 Column 5 Column 6 Column 1 Factor Product Average Duration Shutdown Factor Power At Power To End Of A B %P x A x B Level Power (I P) (Hours) (Hours)
O
- Power Duration = Correction Factor PgxAgxB t=
i
- Must be at least 6 hours Factor A 2 1.5 1 0.7 0.5 0, 3 10 9 8 7 6 5 4 3 l l 1 I I I I I. l l 1 I I I 1 1 I I I l l l l l l t l l l l 1 _
Time From Column 3 (Hours) l f T 1 1 I I I 1 I I i 1 1 I 1
imeFromColumnh(Hoursf I I I I I I I I I I I 1 I I I I I I i I I I I : i I I I I 1 8 9.0 9.3 9.5 ).7 0 1 2 3 4 5 6 7 Factor B O Core Inventory = 1.43E4 X = Curies PCF, col 6 sum Total Corrected Core Activity DATE: TIME: PERFORMED BY:
703445
REVISION PAGE NO.
PROCEDURE NO.
91504-C 0 21 of 21 O
(/
DATA SHEET 18 REACTOR POWER HISTORY AND CORF INVENTORY DETERMINATION NUCLIDE: Pr 144 Column 1 Column 2 Column 3 Column 4 Column 5 Column 6 Average Duration Shutdown Factor Factor Product Power At Power To End Of A B %PxAxB Level Power (I P) (Hours) (Hours)
O
- Power Duration = Correction Factor P xA fxB g=
f i
- Must be at least 1.5 hours Factor A 10 9 8 7 6 5 4 3 2 1.5 1 0.7 0.5 0,3 I I I I l i I I I I I I I I . 1 I I I l 1 1 l l 1 1 1 I I I Time From Column 3 (Hours)I 1 I I I I I I I ' '
O I
'0.2 I O 0.8 '1.0 ' 1. 2 1:4 Time From Column 2 ours) 0.6 I I i 1 1 I I I I I I I I 1
I I I I I I I I I I I I I I I I
(' 1 1 3 4 5 6 7 8 90t 9.3 9.5 6.7 Factor B O Core Inventory = 1.15E4 X = Curies PCF, col 6 sum Total Corrected Core Activity DATE: TIME: PERFORMED BY:
703445
Georgis Pcwsr Company Route 2. B<u 299A Wayrusboro, Georgia 30830 Telephone 404 554 9961 404 724-8114
' Southern Company Services, Inc.
Post Office Box 2625 Birmingharr, Alabama 35202 Telephone 205 870-6011 Vogtle Proj.ect August 23,- 1985 Director of Nuclear Reactor Regulation File: X7BC35 Attention: Ms. Elinor G. Adensam, Chief Log: GN-692 Licensing Branch #4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555 NRC DOCKET NUMBERS 50-424 AND 50-425 CONSTRUCTION PERMIT NUMBERS CPPR-108 AND CPPR-109 V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 SER CONFIRMATORY IIEM 32: PROCEDURE FOR ESTIMATING CORE DAMAGE
Dear Mr. Denton:
Attached for your staff's review are the plant specific procedures for estimating core damage as required by TMI Item II.B.3. This submittal consists of the following procedures:
o Core Damage Assessment;
.o Control Room Output for Assessment of Core Damage; and
.o Core Inventory Determination Using Reactor Power History If your staff requires any additional information, please do not hesitate to contact me.
Sircerely,
- s J. A. Bailey Project Licensing Manager JAB /sm Attachment xc: D. O. Foster- G. Bockhold, Jr R. A. Thomas T. Johnson (W/o Att.)
J. E. Joiner, Esquire D. C. Teper (W/o Att.)
.B. W. Churchill, Esquire L. Fowler M. A. Miller W. C. Ramsey B. Jones, Esquire (W/o Att.) Vogtle Project File L. T. Gucwa 0053V
.