ML13310B126
| ML13310B126 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/02/1983 |
| From: | SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML13310B125 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 PROC-830502, TAC-44478, NUDOCS 8312290295 | |
| Download: ML13310B126 (27) | |
Text
.
g ENCLOSURE 1 Southern California Edison Company Nuclear Engineering & Operations Dept.
REV.4 05/02/83 TABLE 15 EOF PROCEDURE AND CHECKLIST/LOG TECHNICAL GROUP TECHNICAL GROUP DAVE PILMER - LEADER I.
Establish contact with Manager of Operations in TSC.
Unit 1:
PERT #03, PAX 56729 Unit 2/3:
PERT #07, PAX. 56498
- 2.
Establish contact with Technical Support Leader in TSC.
Phone Units 2/3:
56504 Unit 1:
56732 a)
Obtain briefing on plant status.
b)
Determine emergency condition declared.
c)
Determine if offsite assistance has been requested.
d)
Determine mitigating actions underway and cause and prognosis for accident.
- 3.
Establish contact with HSC and provide them with plant status. Phone 572-2300.
- 4.
Determine status of obtaining additional assistance.
- 5.
Brief EOF Liason and ODAC on changing conditions.
- 6.
Provide technical information to the Emergency News Center as directed.
- 7.
Fill in Technical Event Summary (see Attachment 1).
- 8.
Prime responsibility for performing core damage assessment (see Attachment 2).
TABLE IS Page 15.1 8312290295 831227 PDR ADOCK 05000206 P
Southern California Edison Company Nuclear Engineering Operations Dept.
REV. 405702/83 TABLE 15 (Cont'd.)
Attachment I Date:
Timei By:
TECHNICAL EVENT
SUMMARY
Emergency Action Level:
Cause:
Extent of Equipment Failure/Damage:
Compensatory Changes to Plant Equipment/Lineup:
Recovery Efforts in Progress or Planned:
Offsite Assistance Requested?
Personal Injury?
Yes No
_ Name:
==
Description:==
HSC. Advised?
Time By TABLE IS Page 15.2
Southern California Edison Company Nuclear Engineering Operations Dept.
REV. 4 05/02/83 TABLE 15 (Cont'd.)
San Onofre 2 Interim Procedure For Core Damoae Assessment 1.0 PURPOSE A basic procedure is provided for estimating post-accident reactor core damage by using fission product isotopes measured in samples obtained from the Post-Accident Sampling System (PASS).
There are three factors considered in this procedure which are related to the specific activity of the samples. These are the identity of those isotopes which are released.from the core, the respective ratios of the specific activity of those isotopes, and the percent of the source inventory at the time of the accident which is observed to be present in the samples.
The resulting estimate of core damage can be related to one or more of the ten categories described in Enclosure 1. Reference 2.1.contains background information for the basic procedure.
An alternate procedure is provided for situations in which use of the basic procedure is precluded for some reason.
The alternate procedure uses containment radiation measurements as a gross indication of reactor core damage. (Note that the alternate procedure may be used in addition to, as well as instead of, the basic procedure.) Reference 2.2 presents the analysis on which the alternate procedure is based.
2.0.
REFERENCES 2.1 Development of the Interim Procedure Guidelines for Core Damage Assessment, CE Owners Group Task 467, January I 982.
2.2 "Study for Fuel Failure/Fuel Melting and Radiation Monitor Responses for San Onofre Nuclear Generating Station (SONGS) Unit 2," Quadrex Corporation, July 1982.
2.3 50123-111-8.0 "Post-Accident Sampling Program and Analytical Requirements."
3.0 DEFINITIONS 3.1 Fuel Damage: For the purpose of this procedure, fuel damage is defined as a progressive failure of the, material boundary to prevent the release of radioactive fission products into the reactor coolant starting with a penetration in the zircaloy cladding. The type of fuel damage as determined by this procedure is reported in terms of four major categories which are: no damage, cladding failure, fuel overheat, and fuel melt.
Each of these categories are characterized by the identity of the fission products released, the mechanism by which they are released, and the source inventory within the fuel rod from which they are released. The degree of fuel damage is measured by the percent of the fission produce source inventory which has been released into fluid media and therefore available for immediate release to the environment. The degree of fuel damage as determined by this procedure is reported in terms of three levels which are: initial, intermediate, and major. This results in a total of ten possible categories as characterized in Enclosure I 3.2 Source Inventory: The source inventory is the total quantity of fission products expressed in curies of each isotope present in either source; the fuel pellets or the fuel rod'gas gap.
4.0 PRECAUTIONS AND LIMITATIONS 4.1 The assessment of core damage obtained by using this procedure is only an estimate. The techniques employed in this procedure are adequate only to locate the core condition within one or more of the 10 categories of core damage described in Enclosure I'. The procedure is based on radiological data.
Other plant indications may be available which can improve upon estimation of core damage. These include incore temperature indicators,^the total quantity of hydrogen released from zirconium degradation and containment radiation monitors. Whenever possible, these additional indicators should be factored into the assessment.
4.2 This procedure relies upon samples taken from multiple locations inside the containment building to determine the to'tal quantity of fission products available for release to the environment. The amount of-fission products present at each sample location may be changing rapidly-due to. transient plant conditions. Therefore, it is required that the samples should be obtained-within a minimum time period and if possible. under stabilized plant conditions.
Samples obtained during rapidly. changing plant conditions should not be weighed heavily into the assessment of core damage.
TABLE 15 Page 15.3
Southern California Edison Company Nuclear Engineering & Operations Dept.
REV. 4 05/02/83 TABLE 15 (Cont'd.) (Cont'd.)
4.3 A number.of factors influence the reliability of the chemistry samples upon which this procedure is based.
Reliability is influenced by the ability to obtain representative samples due to incomplete mixing of the fluids, equipment limitations, and lack of operator familiarity with rarely used procedures.
The accuracy achieved in. the radiological analyses are also influenced by a number of factors. The equipment employed in the analysis may be subjected to high levels of radiation exposure over extended periods of time. Chemists are required to exercise considerable caution to minimize the spread of radioactive materials.
Samples have the potential of being contaminated by numerous sources and they may not. represent the average distribution of the contaminants in the sampled fluid.
Cooling or reactions may take ploce-in the long sample lines. Therefore, the results obtained may not be representative of plant conditions. To minimize these effects, multiple samples should be obtained over an extended time period from each location.
5.0 PLANT CONDITION/SYMPTOMS This procedure is to be employed.for analysis of radiochemistry sample data when it is determined that a plant accident with the potential for core dmage has occurred. The following is a list of plant symptoms to assist in this determination. This list is not a complete-representation of all events or conditions which may indicate potential core damage. However, the existence of one or more of these events or conditions signals a possible need to activate this procedure.
5.1 High alarm on the containment radiation monitor.
5.2 High alarm on the CVCS letdown radiation monitor.
V 5.3 High alarm on the main condenser air ejector exhaust radiation monitor.
5.4 Pressurizer level low.
5.5 Safety Injection System may have automatically actuated.
5.6 Possible high quench tank level, temperature, or pressure.
5.7 Possible noise indicative of a high energy line break.
5.8 Decrease in volume control tank level.
5.9 Standby charging pumps energized.
5.10 Unbalanced charging and letdown flow.
5.11 Reactor Coolant System subcooling low or zero.
6.0 PREREQUISITES An operational Post-Accident Sampling System with the capability to obtain and analyze the identity and concentration of fission product isotopes in accordance with the provisions of Reference 2.3.
7.0 BASIC PROCEDURE 7.1 Record of Plant Condition Record the following plant indications using Enclosure 2 as a worksheet. The values should be recorded as close as possible to the time at which the radiological samples are obtained from the Post-Accident Sampling System.
If additional samples are taken at a later time, record another set of values at or near that time.
TABLE 15 Page 15.4
01 Southern California Edison. Company Nuclear Engineering & Operations Dept.
REV. 405/02/83 TABLE 15 (Cont'd.) -(Cont'd.)
7.1.1 Reactor Coolant System:
Pressure Temperature Reactor Vessel Level Pressurizer Level.
7.1.2 Containment Building:
Atmosphere Pressure Atmosphere Temperature Sump Level 7.1.3 Prior 30 Days Power History 7.1.4 Time of Reactor Shutdown 7.2 Selection of Samole Location Obtain specific activity data from samples of the reactor coolant, the containment sump water, and the containment atmosphere.
7.3 Sample Recording Record the required data for each sample. Enclosure 3 is provided as a worksheet.
Some of the isotopes listed in the enclosure may not be observed in the sample.
7.4 Temperature and Pressure Correction Correct the measured sample specific activ~ty to standard temperature and pressure.
7.4.1 Reactor coolant liquid samples are corrected for temperature using the factor for wafer density from Enclosure 4.
The measured value of specific activity is divided by the correction factor corresponding to the sample temperature from Enclosure 3.
This corrected value of specific activity is recorded in Enclosure 5.
7.4.2 Containment building sump samples <b not require correction for temperature and pressure within the accuracy of this procedure.
7.4.3 Containment building atmosphere son ples are corrected for temperature and pressure using the following equation.
S47T 1
+ 460 14.7 46 Specific Activity (STP)= Specific Act vity x (
1 XM) x(
where:
P Measured saple pressure, sig, from Enclosure 3 T= Measured sample temperat re, Of, from Enclosure 3 Record the correct d values of speciflc activity on Enclosure 5.
TABLE 1 ge 15.5
Southern California Edison Company Nuclear Engineering & Operations Dept.
REV. 4 05/02/83 TABLE I5 (Cont'd.) (Cont'd.)
7.5 Decay Correction Correct the sample specific activity for decay back to the time of reactor shutdown using the following equation. Enclosure 6 is provided as a worksheet.
A0 =
A e -Xt where:
Ao =the specific activity of the sample corrected back to the time of reactor shutdown, ci/cc.
A = the measured specific activity, Mci/cc.
= the radioactive decay constant, I/sec.
t = the time period from reactor shutdown to sample analysis, sec.
7.6 Identification of the Fission Product Release Source 7.6.1 Calculate the following ratios for each noble gas isotope and each iodine isotope using the decay corrected specific activities recorded on Enclosure 6. Enclosure 7 is provided as a worksheet.
Noble Gas Ratio = Noble Gas Isotope Specific Activity Xe-133 Specific Activity Iodine Ratio = Iodine Isotope Specific Activity 1-131 Specific Activity 7.6.2 Determine the source of release by comparing the results obtained to the predicted ratios provided in Enclosure 7. Identify as the source that ratio.which is closest to the value, obtained in step 7.6.1.
7.7 Quantitative Release Assessment Calculate the total quantity of fission products available for release-to the'environment. Enclosure 8 is provided as a worksheet..
7.7.1 The quantity of fission products in the reactor coolant is determined as follows:
7.7.1.1 If the water level in the reactor vessel recorded on Enclosure 2 indicates that the vessel is full, the quantity of fission products in the reactor coolant is calculated by the following equation.
Total Activity (Ci) = A ( -Mc/cc) x RCS Volu me (STP) x 0-6 ci /
C where:
A0 = the specific activity of the reactor coolant sample recorded on,
FcI/cc.
RCS Volume (STP) = the full reactor coolant system water volume corrected to standard temperature by-multiplying by the Jactor for water density
-provided in Enclosure 4.
TABLE IS Page 15.6
Southern California Edison Company Nuclear Engineering Operations Dept.
REV. 4 05/02/83 TABLE 15 (Cont'd.) (Cont'd.)
7.7.1.2 If the water levels in the reactor vessel recorded on Enclosure 2 indicates that a steam void is present in the reactor vessel,. then the quantity of fission products found in the reactor coolant is again calculated by step 7.7.1.1.
However, it must be recognized that the value obtained will overestimate the actual quantity released. Therefore, this sample should be repeated at such time when the plant operators have removed the void from the reactor vessel.
7.7.1.3 If the water level in the reactor vessel recorded on Enclosure 2 is below the low end capability of the indicator, it is not possible to determine the quantity of fission products from this sample because the volume of water in the reactor coolant system is unknown.. Under this condition, assessment of core damage is obtained using the containment sump sample.
7.7.2 The quantity of fission products in the containment building sump is determined as follows:
7.7.2.1 The water volume in the containment building sump is determined from the sump level recorded on Enclosure 2 and the curve provided in Enclosure 9.
(Note If indicated level is less than 18 feet,. use a sump water volume. of 7.7.2.2 The quantity of fission products in the sump is calculated by the following equation.
Total activity, Ci = Ao Ic'/cc) x Sump Water Volume x 10-c /
Ci where:
Ao = the specific Gctivitr of the containment sump sample recorded on,
Fc /cC.
7.7.3 The quantity of fission products found in the containment building atmosphere is determined, as follows:
7.7.3.1 The volume of gas in the containment building is corrected to standard temperature and pressure using the following equation.
Pl+
)4. (
492 Gas Volume (STP) = Gas Volume x ---
where:
T= Containment atmosphere temperature, OF, from on Enclosure 2.
P, = Containment atmosphere pressure, psig, from Enclosure 2.
7.7.3.2 The quantity of fission products in the containment is calculated by the following equation.
Total Activity, Ci Ao.
c/cc)
Gas Volume (STP) x 10-6 Ci/
Ci where:
A0 = The specific activity of the containmenit building sample recorded on,
M ci/C.
7.7.4 The total quantity of fission products avoilable for release to the environment is equal to the sum of the values obtained from each sample location Record th total quantity of fission products on Enclosure 8.
TABLE 15 Page 5.
Southern California Edison Company Nuclear Engineering & Operations Dept.
REV. 4 05/02/83 TABLE 15 (Cont'd.) (Cont'd.)
7.8 Plant Power Correction The quantitative release of the fission products is expressed as the percent of the source inventory at the time of the accident.
The equilibrium source inventories are to be corrected for plant power history.
7.8.1 To correct the source inventory for the case in which plant power level has remained relatively constant prior-to reactor shutdown the following procedure is employed. 0 is provided as a worksheet.,
7.8.1.1 The fission products are divided into.two groups based upon the radioactive half lives. Group I isotopes ore to be employed in the case where core power had not changed more than +10 percent within the last 30 days prior to the reactor shutdown.
Group 2 isotopes are to be employed in the case where core power hod not changed more than +10 percent within the last 4 days prior to the reactor shutdown.
7.8.1.2 The following equation is applied to the appropriate fission product group.
Group I Power Correction Factor =Averace Power for Prior 30 Days, %
100 Group 2 Power Correction Factor = Averace Power for Prior 4 Days, %
100 NOTE: Use Step 7.8.1 only if sampling has detected the presence of at least two isotopes in the Group to be employed.
Otherwise, use Step 7.8.2 7.8.2 To correct the source inventory for the case in which plant power level has not remained constant prior to reactor shutdown, the following equation is employed. The entire 30 days.
power history should be employed. Enclosure II is provided as a worksheet.
IjP (1e - Atj)e-t Power Correction Factor =
100 where:
P= steady reactor power in period j, t=
duration of period j, sec.
0 t = time from end of period j to reactor shutdown, sec.
A = decay constant for isotope, sec 7.9 Comparison of Measured Data with Source Inventory The total quantity of fission products available for release to the environment recorded in Enclosure 8 is compared to the source inventory corrected for plant power history recorded in Enclosure 10 or II.
This comparison is made by dividing the two values for each isotope and calculating the percent of the corrected source inventory that is now in the sampled fluid and therefore available for release to the environment. Enclosure 12 is provided as a worksheet.
TABLE 15 Page 15.8
Southern California Edison Company Nuclear Engineering & Operations Dept.
REV.4 05/02/83 TABLE 15 (Cont'd.) (Cont'd.)
7.10 Assessment of Core Damage The conclusion on core damage is made using the three parameters developed above. These are:
- 1.
Identification of the fission product isotopes which most characterize a given sample (Enclosure 3).
- 2.
Identification of the source of the release (Enclosure 7).
- 3.
Quantity of the fission produce available for release to the environment expressed as a percent of source inventory (Enclosure 12).
Knowledgeable judgment is used to relate the above three parameters to the definitions of the 10 categories of fuel damage found in Enclosure 1.
Core damage is not anticipated to take place uniformly. Therefore, when evaluating the three parameters listed above the procedure is anticipated to yield a combination of one or more of the 10 categories defined in Enclosure I. These categories will exist simultaneously.
8.0 ALTERNATE PROCEDURE The alternate core damage assessment procedure is based on calculated values of containment radiation levels in the period following an unmitigated large break LOCA. These calculations considered the release of fission products from gap and fuel, their escape from the RCS, and their subsequent dispersal within the containment building.
The phenomena treated in such an analysis are difficult to model and a number of simplifying assumptions must be mode.
The results, therefore, are subject to large uncertainties.
Thus, the alternate procedure produces only a rough estimate of the degree of core damage.
8.1 Limitations and Precautions 8.1.1 The alternate procedure is based on calculations of containment radiation levels ofter-a, double-ended cold leg break, with no post-accident cooling of the core.
Use of the'.
procedure following an occident of lesser severity will tend to underestimate the extent of core damage.
8.1.2 The calculations on which the alternate procedure is based assumed full-power, equilibrium values of core fission product inventories. Use of the procedure in a situation in which the reactor has been operating for a short time and/or at lower power will tend to underestimate the extent of core damage.
8.2 Application of Procedures 8.2.1 Record readings of dose rate (mr/hr) from both of the containment high-range radiation monitors (2RE-7802-1 and 2RE-7820-2). Take the averoge of these two values.
8.2.2 Estimate the elapsed time (hours) between the initial release of fission products from the core and the time when dose rate readings are taken. (If a sudden, pronounced increase in dose rate is observed, the new, higher dose rate should be recorded immediately. In this case, the elapsed time is considered to be zero.)
8.2.3 On Enclosure 13, record the point which corresponds-to the averaged value of dose rate from 8.2.1 and the elapsed time from 8.2.2.
8.2.4 Interpolate between the curves of Enclosure 13 to estimate the percentage of clad ruptures or of melted fuel represented by the point recorded in 8.2.3.
8.2.5 Repeat steps 8.2.1 through 8.2.4 subsequently as freqtently as practicable.
TABLE IS Page 15.9
cjh rn Radiological Characteristics of NRC Categories of Fuel Damage Release of Characteristic NRC Category of Mechanism of Source of Characteristics Isotope Expressed as a Fuel Damage Release Release Isotope Percent of Source Inventory
- 1.
No Fuel Damage Haloogen Spiking Gas Gap I 131, Cs 137 Less than I Tramp Uranium Rb 88
- 2.
Initial Cladding Gas Gap Less than 10 Failure.
- 3.
Intermediate Clad Burst and Gas Gap Xe 131m, Xe 133 10 to 50 0
Cladding Failure Gas Gap Diffusion 1 131, 1 133 O
Release
- 4.
Major Cladding Gas Gap Greater than 50 0
Failure ct a]
S.
Initial Fuel Pellet Fuel Pellet Cs 134, Rb 88, Less than 10 r V Overheating Te 129, Te 132 Intermediate Grain Boundary Fuel Pellet 10 to 50
()
O Fuel Pellet Diffusion S
(I)
Overheating E,.
- 7.
Major Fuel Pellet Diffusional Release Fuel Pellet Greater than 50 Overheating From U0 2 Grains f
Initial Fuel Pellet Fuel Pellet Less than 10 Melt Intermediate Fuel Escape from Molten Fuel Pellet Ba 140, La 140 10 to 50 Pellet Melt Fuel La 142, Pr 144 Major Fuel Pellet Fuel Pellet Greater than 50 Melt A..
C) 3.o
Southern California Edison Company Nuclear Engineering & Operations Dept.
REV. 405/02/83 TABLE IS (Cont'd.) (Cont'd.)
ENCLOSURE 2 Record of Plant Indications Dote Time Reactor Coolant System Pressure
._psig Temperature OF Vessel Level Pressurizer Level RCS Volume 3.14 x 108 cc Containment Building Pressure psig Temperature OF Sump Level Ft Containment Volume 6.6 x 10 10 cc Prior 30 Days Power History Power
%for days, then Power
%for days, then Power
%for 'v days, then Power
%for..
days to time of shutdown Reactor Shutdown Time Date TABLE 15 Page 5. II
Southern California Edison Company Nuclear Engineering & Operations Dept.
REV. 4 05/02/83 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 3 Record of Sample Specific Activity Sample Number:
Location:
Time of Analysis:
Temperature, OF:
Pressure, PSIG:
Isotope Sample Activity, uc'/cc:
Kr 87 Xe 131m Xe 133 1 131 1 132 I 133 1 135 Cs 134 Rb 88 Te 129 Te 132 Sr 89 Ba 140 La 140 La 142 Pr 144 TABLEI5.
15-aeI
0 Southern California Edison Company Nuclear Engineering & Operations Dept.
REV. 4 05/02/83 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 4 DENSITY CORRECTION FACTOR FOR REACTOR COOLANT TEMPERATURE ENCLOSURE 4 Density Correction Factor for Reactor Coolant Temperature T, oF 800 700 600 500 400 300 200 100 0
0 0.25 0.50 0.75 1.00 1.25 Factor TABLE 15 Page 15.13
Southern Californa Edison Company Nuclear Engineering & Operations Dept.
REV. 4 05/02/83 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 5 Record of Sample Temperature and Pressure Correction Sample Number:
Location:
Time of Analysis:
Temperature, OF:
Pressure, PSIG:
Measured Specific Activity Correction Specific Activity Isotope (Enclosure 3), 'c/cc Factor at STP, 9c1 /cc Kr87 Xe 131m Xe 133 1131 1 132 1133 1135 Cs 134 Rb 88 Te 129 Te 132 Sr 89 Ba 140 La 140 La 142 Pr 144
'TABLE 15 F ae 151.4
Southern California Edison Company Nuclear Engineering & Operations Dept.
REV. 4 05/02/83 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 6 Record of Samole Decay Correction Sample Number:
Location:
Time of Reactor Shutdown:
Time of Analysis:
Elasped Time, t:
Decay Specific Activity Decay Corrected
- Constant, at STP (Enclosure 5),
Specific Activity, A
fA)
(A)*
Isotooe I/sec c
ecc Kr 87 1.5 (-4)
Xe 131m 6.7 (-7)
Xe 133 1.5 (-6) 1 131 9.9(-7) 1 132 8.4 (-5)
I 133 9.3 (-6) 1 135 2.9(-5)
Cs 134 1.1 (-8)
Rb 88 6.5(-4)
Te 129 1.7 (-4)
Te 132 2.5 (-6)
Sr 89 1.61(-7)
Ba 140 6.3 (-7)
La 140 4.8 (-6)
La 142 1.2 (-4)
Pr 144 6.7 (-4)
- ^
o A
0 TABLE15 Pagel15.1
Z C
0 Record of Fission Product Release Source Identification r)
Sample Number:
Location:
(0 Decoy Corrected Specific Activity Calculated Activity Ratio Activity Ratio Identified Isotope (Enclosure 6), p'/cc Isotope Ratio*
in Fuel Pellet In Gas Gap Source (O
Kr 87 0.2 0.001 Xe 131m 0.003 0.001 0
~0 Xe 133 1.0 1.0 1.0 1131 1.0 1.0 1.0 m
eta z9
-o 1132 1.4 0.01 O
2.0 0.5 1135 1.8 0.17 O
Noble Gas Ratio - Decay Corrected Noble Gas Specific Activity Decay Corrected Xe 133 Specific Activity ine Ratio
= Decay Corrected lodine Isotope Specific Activity Decay Corrected I 131 Specific Activity lj C3 JI:
- 4.
PCO
Southern California Edison Company Nuclear Engineering & Operations Dept.
REV. 4 05/02/83 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 8 Record of Release Quantity Reactor Coolant Containment Sump Containment Total Sample Number, Sample Number, Atmosphere Sample Quantity IsotoDe Ci Ci Number, Ci Ci Kr87 Xe 13Im Xe 133 1131 1132 1133 1135 Cs 134 Rb 88 Te 129 Te 132 Sr 89 Ba 140 La 140 La 142 Pr 144 TABLE 15 Page 15.17
Southern California Edison Company Nuclear Engineering Operations Dept.
TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 9 CONTAINMENT BUILDING SUMP LEVEL ENCLOSURE 9 Containment Building Sump Level Level, Ft.
26 25 24 23 22 21 20 19 18 0
4 8
12 16 20 24 28 Volume, cc x 10 TABLE I5 Page 1518
Southern California Edison Company Nuclear Engineering Operations Dept.
REV. 405/02/83 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 10 Record of Correction for Constant Power Level Sample Number:
Location:
Average Power for Prior 30 Days:
Average Power for Prior 4 Days:
Fuel Power Equilibrium Corrected History Correction Source Source Isotope Group Factor, F Inventory, Ci Inventory. Ci Gas Goo Inventory Kr 87 2
9.5(0)
Xe 131m I
6.6(4)
Xe 133 I
1.8(7) 1 131 1
9.0 (6) 1 132 2
9.9 (3) 1 133 2
8.9 (6) 1 135 2
1.6 (6)
Fuel Pellet Inventory Kr 87 2
4.7 (7)
Xe 131m I
7.0(5)
Xe 133 2.0 (8) 1 131 9.9 (7) 1 132 2
1.4(8) 1 133 2
2.0(8).
1 135 2
1.9(8)
Cs 134 1
1.8 (7)
Rb 88 2
6.8 (7)
Te 129 2
3.1 (7)
Te 132 I
1.4 (8)
Sr 89 I
9.4.(7)
Ba 140 I
1.7(8)
La 140 I
1.8(8)
La 142 2
2.2 (8)
Pr 144 2
1.2 (8)
F = Average power for prior 30 days, %
(Group I)
F = Average power for prior 4 days, %
(Group 2)
TABLE IS Page 15.19
Southern California Edison Company Nuclear Engineering Operations Dept.
REV. 4 05/02/83 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE I Record of Correction for Nonconstant Power Level (See Enclosure 2 for Power History)
Decay Power Equilibrium Corrected Constant Correction Source Source IsotoDe (0) 1/sec Foctor, F Inventory, Ci nventory C Gas GaD
-nventory, Kr 87 1.5 (-4) 9.5 (0)
Xe 131m 6.7 (-7) 6.6 (4)
Xe 133 1.5 (-6) 1.8 (7) 1 131 9.9 (-7) 9.0 (6) 1 132 8.4 (-5) 9.9 (3) 1 133 9.3 (-6) 8.9 (6) 1 135 2.9 (-5) 1.6 (6)
Fuel Pellet Inventory Kr 87 1.5 (-4) 4.7(7)
Xe 131m 6.7 (-7) 7.0 (5)
Xe 133 1.5 (-6) 2.0 (8) 1 131 9.9 (7) 9.9 (7) 1 132 8.4 (-5) 1.4 (8) 1 133 9.3(-6) 2.0(8) 1 135 2.9 (-5) 1.9 (8)
Cs 134 1.1 (-8) 1.8(7)
Rb 88 6.5(-4) 6.8(7)
Te 129 1.7 (-4) 3.1 (7)
Te 132 2.5(-6) 1.4 (8)
Sr 89 1.6 (-7) 9.4 (7)
Ba 140 6.3 (-7) 1.7 (8)
La 140 4.8 (-6) 1.8(8)
La 142 1.2(-4) 2.2 (8)
Pr 144 6.7(-4) 1.2,(8)
.-At.
-At.
.P.
(I-e
) e F
100 TABLE 15 Page 15.20
Southern California Edison Company Nuclear Engineering & Operations Dept.
REV. 4 05/02/83 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 12 Record of Percent Release Total Quantity Power Corrected Available for Release Source Inventory, Isotope (Enclosure 8), Ci Ci (Enclosure 10 or 11)
Percent Gas Ga Inventory Kr 87 Xe 131m Xe 133 1131 1 132 1 133 1 135 Fuel Pellet Inventory Kr 87 Xe 131m Xe 133 1131 1 132 1 133 1 135 Cs 134 Rb 88 Te 129 Te 132 Sr 89 Ba 140 La 140 La 142 Pr 144 TABLE IS P
e TAB LE 15
-Page-1-21.
Southern California Edison Company Nuclear Engineering & Operations Dept.
REV. 4 05/02/83 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 13 Containment Monitor Dose Rate vs. Time After Initial Release for Various Percentages of Clad Failure or Core Melt Dosage MR/HR 109
-- 100%
Fuel 10 10%
Melt 107 100%
10 30%
Clad
.10%
Failure 1%
10 0
1 2.345678 Tie'(Hours)
TABLE 15 ag 522
ENCLOSURE 2 High Radiation Sampling System Influent Chemistry Design Parameters Reactor Coolant Containment Atm cc(@STP)
H2 0-2000 Kg
- 0-10 Volume%
02 0-20 ppm Boron 100-4400 ppm pH 3-12 Cl-
- 0-20 ppm Radio Chemistry:
Nuclides (gross) 1-107 uCi/cc 10-2-105 uci/cc The containment hydrogen concentration is not measured using the PASS, but will be monitored using the containment hydrogen monitor.
- The chloride monitoring capability will be provided by the undiluted sampling capability mentioned in the NRC's order dated March 14, 1983.
LABennett:8875
Eric I osuFr< 3 ;
RESULTS OF PASS DEMONSTRATION CONDUCTED APRIL 5, 1983 MEASURED PARAMETER PASS CHEM LAB DEVIATION TOLERANCE RCS pH 5.9 5.89 O.OlpH Unit
+0.3pH Unit RCS Boron 1850ppm 1755ppm
+ 5.4%
+5%
RCS Total Gas 170cc/kg 188cc/kg
-9.6%
+20%
RCS 02 0.15%*
0.0%
0*
None Specified RCS H2 15.28 cc/kg**
16.0 cc/kg 0.72 cc/kg**
+5cc/kg RCS Radio-Calibration N/A N/A N/A nuclides Source Used Containment 0.0%
0.0%
0
+20%
H2 Containment Calibration N/A N/A N/A Radionuclides Source Used
- Data listed is from second sample, 0.15% is the amount of 02 present in the nuclear service water used for flushing.
- Data listed is from second sample. The most likely cause of the discrepancy in the H2 analysis is the time difference between PASS and Chem.
Lab samples (approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) and the probability that equilibrium H2 conditions did not exist in the RCS.
kI SURVEILLANCE CRITERIA The attached table provides the specific surveillance criteria proposed by SCE for implementation of the NRC requirements on accuracies of the instrumentation employed in the Post-Accident Sampling System.
The surveillance criteria stated in this table agree directly with the NRC guidance for instrument accuracies.. A basis is stated for the case of each instrument minimum sensitivity. The subject surveillance criteria will be implemented into the definition of Plant Accident Sampling System through established station procedures.
-2 PASS Proposed Function Surveillance Criteria Basis pH
+ 0.3 pH units between pH Defined by NRC guidance.
75to 9
+ 0.5 pH units in all other ranges Boron
+5 percent of the comparative The range for boron measurement Taboratory measurement -but no is 0 to 6000 ppm.
The less than + 100 ppm.
important value is the boron concentration required for reactor shutdown which is about 1800 ppm.
Five percent of this value is approximately 100 ppm. Therefore, the value 100 ppm was selected as the minimum sensitivity and is consistent with steady state boron concentrations under normal operations.
Total
+ 20 percent of the comparative The range for this measurement is dissolved Taboratory measurement but no 0 to 2000 cc/kg. A mid-range gas less than + 100 cc/kg.
value which may be anticipated under post accident degraded core conditions is 500 cc/kg. Twenty percent of this value is 100 cc/kg. Therefore, the value 100 cc/kg is employed as the minimum sensitivity.
The value of 100 cc/kg is consistent with values of total gas found under normal operating conditions. It is not anticipated that the plant will operate below 50 cc/kg.
Oxygen No surveillance required This instrument is recommended but gas not required by NUREG-0737. No surveillance will be implemented in terms of the definition of PASS operability. However, standard maintenance practices will be employed to keep this instrument operable.
+ 20 percent of the comparative The range of this measurement is gas Taboratory measurement but no 0 to 2000 cc/kg which is less than a minimum sensitivity 100 percent of the range of of 25 cc/kg.
total gas. The minimum sensitivity of 25 cc/kg is employed because this value is the minimum required for RCS hydrogen concentration under normal plant operations. - It is not anticipated that the plant will operate below this value.
Containment
+ 20 percent of the comparative No specific NRC criterion is Atmosphere Taboratory measurement but no defined for the accurary of the Hydrogen less than a minimum sensitivity containment hydrogen measurement.
of 1 percent hydrogen The accuracy stated is.consistent concentration.
with the reactor coolant hydrogen measurement. The minimum sensitivity is below the lower limit for flamability of a hydrogen gas mixture in air.
Gross Accurate within a factor of two Defined by NRC guidance.
Activity across the entire range of Gamma IC1/ml to 10 Ci /ml.
Spectrum