ML13324A675
| ML13324A675 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 07/17/1985 |
| From: | Medford M SOUTHERN CALIFORNIA EDISON CO. |
| To: | Zwolinski J Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM TAC-44478, NUDOCS 8507190210 | |
| Download: ML13324A675 (31) | |
Text
Southern California Ec'ison Company P. 0. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD. CALIFORNIA 91770 M. 0. MEDFORD TELEPHONE MANAGER, NUCLEAR LICENSING July 17, 1985 (818) 302-1749 Director, Office of Nuclear Reactor Regulation Attention: J. A. Zwolinski, Chief Operating Reactors Branch No. 5 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:
Subject:
Docket No. 50-206 NUREG-0737, Item II.B.3 - Post-Accident Sampling Capability San Onofre Nuclear Generating Station Unit 1
References:
- 1. Letter, D. M. Crutchfield, NRC, to R. Dietch, SCE, TMI Action Plan Item II.B.3, Post-Accident Sampling System, September 1, 1983
- 2. Letter, M. 0. Medford, SCE, to D. M. Crutchfield, NRC, Post-Accident Sampling Capability, December 27, 1983
- 3. Letter, R. E. Uhrig, FP&L, to D. G. Eisenhut, NRC, Report on PASS Valve Operability and Instrumentation Testing, St. Lucie Unit No. 2, Docket No. 50-389, December 29, 1982 Reference 1 provided us with the NRC staff's evaluation of our December 3, 1982 submittal regarding the post-accident sampling system (PASS) for San Onofre Unit 1. The evaluation found that we fully met nine of the eleven criteria in Item II.B.3 of NUREG-0737. The evaluation discussed the two open items (Criterion 2 and Criterion 10) and requested that we address the open items in order to enable the NRC staff to complete their review.
Accordingly, Reference 2 provided you with our responses to the open items and indicated that it should satisfy the staff's request. However, subsequent discussions with the staff as part of the PASS licensing actions for the return-to-service in November 1984, indicated that additional information was required and that it should be provided prior to final PASS implementation per Provisional Operating License No. DPR-13, License Condition 3.K, "Post-Accident Sampling System (PASS), NUREG-0737, Item II.B.3."
Accordingly, this letter provides the additional information to resolve these open items.
The information is formatted to document each open item and our response.
OPEN ITEM 1 The NRC staff finds that the licensee partially meets Criterion (2) by establishing an on-site radiological and chemical analysis capability.
However, the licensee should provide a procedure, consistent with our clarification of NUREG-0737, Item II.B.3, Post Accident Sampling System, transmitted to the licensee on June 30, 1982, to estimate the extent of core 8507190210 850717 PDR ADOCK 05000206 P
-PDR,
Mr. 3.
July 17, 1985 damage based on radionuclide concentrations and taking into consideration other physical parameters such as core temperature data and sample location.
Guidance for the procedure to estimate core damage is attached (Attachment 1).
In addition, the licensee should describe the accuracy of the chloride analysis.
SCE RESPONSE TO ITEM 1 The procedure for core damage assessment at San Onofre Unit 1, provided as to this letter, was developed prior to return to power from the seismic backfit outage which ended on November 27, 1984. The procedure incorporates the Westinghouse Owner's Group Core Damage Assessment guidance as it applies to San Onofre Unit 1. The post-accident assessment of core condition using this procedure will be performed by the technical support group in the Emergency Operations Facility.
The chloride analysis will be performed as part of the undiluted grab sample analysis performed for SCE by General Atomic Technologies Co. at their San Diego hot cell facility. The accuracy of the chloride analysis will be consistent with the clarification to Criterion 10 provided in the NRC letter dated June 30, 1982, for concentrations between 0.5 and 20.0 ppm chloride for which the accuracy should be + 10%. However below 0.5 ppm the instrumenta tion cannot detect chlorides with any quantifiable accuracy. The lack of accuracy at this extremely low level is not considered significant.
OPEN ITEM 2 The NRC staff finds that the licensee partially meets Criterion (10).
The licensee should verify that the accuracy is consistent with the guidelines in our letter dated June 30, 1982 for chloride analysis.
The licensee should also provide information on the measurement ranges and sensitivity of the procedure to demonstrate on the standard test matrix that the selected procedures and instrumentation will achieve the accuracies.
SCE RESPONSE TO ITEM 2 The accuracy of the chloride analysis is discussed in our response to Open Item 1 above.
In order to demonstrate the operability of the San Onofre Unit 1 Combustion Engineering (C-E) PASS components and instrumentation in the post-accident water chemistry and radiation environment, C-E has previously prepared a report entitled "Engineering Evaluation and Functional Testing for the CE PASS Components and Instrumentation," November 1982, for St. Lucie Unit No. 2, Docket No. 50-389. Reference 3 submitted this report for NRC staff review.
This report was reviewed by the NRC staff and found to provide verification of the adequacy of the St. Lucle 2 PASS under the anticipated water chemistry and radiation environment. The acceptance of the C-E report
Mr. J. A. Zwolinski
-3_
July 17, 1985 is documented in NUREG-0843, Supplement No. 3, "Safety Evaluation Report related to the operation of St. Lucie Plant, Unit No. 2, Docket No. 50-389,"
dated April 1983.
SCE has reviewed the above discussed report and determined that all of the components and instrumentation of the St. Lucie 2 C-E PASS are identical to the SONGS 1 C-E PASS. The anticipated SONGS-1 post-accident chemistry and radiation environment is not significantly different than that which would be expected at St. Lucie 2 and, accordingly, the report results are applicable to the SONGS-1 PASS. Therefore, it is concluded that the instrumentation and components of the San Onofre Unit 1 PASS are appropriate for use in the expected San Onofre Unit 1 post-accident water chemistry and radiation environment and the San Onofre Unit 1 PASS meets the requirements of Criterion
- 10.
The above discussed and enclosed information should resolve these NUREG-0737, Item II.B.3 open items. If you have need for any additional information to complete your review, please let us know.
Very truly yours, Enclosure cc:
- 3. B. Martin, Regional Administrator, NRC Region V
ENCLOSURE 1
0 0
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 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 I).
- 8.
Prime responsibility for performing core damage assessment (see Attachments 2, 3145, and 6.
TABLE IS Page 15.1
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.)
Attachment I Date:
Time:
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 15 Page 15.2
Attachments 2, 3, and 4, Pages 15.3 -
15.42 Pertain to San Onofre Units 2 and 3
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.)
San Onofre Unit I Procedure For Core Damage Assessment Using Radiological Samples 1.0 PURPOSE This basic procedure is employed for estimating post-accident reactor core damage by sampling for fission product isotopes.
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 I. Reference 2.1 contains background information for the basic procedure.
A computer program for performing the basic procedure calculations is described in Reference 2.2.
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.) The alternate procedure is based on information presented in Reference 2.3.
2.0 REFERENCES
2.1 Development of the Comprehensive Procedure Guidelines for Core Damage Assessment, CE Owners Group Task 467, July 1983.
2.2 Program Documentation for "P094, Assessment of San Onofre Core Damage", SCE D/P-Engineering Programming.
2.3 Letter, L. D. Brevig to G. Gibson, "Alternate Methods of Unit I Post-Accident Parameter Sampling,"
April 9, 1984.
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 PRECAULTONS 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. This procedure is based on radiological data.
Other plant indications may be available which can improve upon estimation of core damage. These include the total quantity of hydrogen released from zirconium degradation and containment radiation monitors. Whenever possible, these additional indicators should be factored into the assessment.
TABLE 15 Page 15.43
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15185 TABLE 15 (Cont'd.) (Cont'd.)
4.2 This procedure relies upon samples taken from multiple locations inside the containment building to determine the total 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.
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 lock 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 place 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 damage 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.
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 The capability to obtain samples of reactor coolant, containment sump liquid, and containment atmosphere and to analyze these to determine the concentration of fission product isotopes.
TABLE 15 Page 15.44
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
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. If additional samples are taken at a later time, record another set of values at or near that time.
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 Sample 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 activity to standard temperature and pressure.
7.4.1 Reactor coolant liquid samples are corrected for temperature using the factor for water 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 do not require correction for temperature and pressure within the accuracy of this procedure.
TABLE 15 Page 15.45
Southern California Edison Company Nuclear Engineering, Safety & Licensing Oept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
7.4.3 Containment building atmosphere samples are corrected for temperature and pressure using the following equation.
1 *2
+ 460 Specific Activity (STP)= Specific Activity x (P + 14.7 x
492 where:
P = Measured sample pressure, psig, from Enclosure 3 T = Measured sample temperature, OF, from Enclosure 3 Record the corrected values of specific activity on Enclosure 5.
7.5 Adjustments to Scmple Specific Activity 7.5.1 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.
Ao _
A
-At where:
A0 = the specific.activity of the sample corrected back to the time of reactor
- shutdown, ci/cc.
A
= the measured specific activity, pci/cc.
= the radioactive decay constant, I/sec.
t
= the time period from reactor shutdown to sample analysis, sec.
7.5.2 Correction for Activity Produced by Precursors For certain isotopes, the total specific activity includes that formed by decay of a precursor or precursors.
For these isotopes it is necessary to remove the precursor contribution to total activity, so that the remainder represents only the isotope itself as a fission product.
The decay corrected specific activity, A0, is multiplied by a fractional factor which yields the portion of the activity attributable to the isotope itself, A'o. Thus:
A'o =f Ao and e-At f
=
ek Apt -e
+eAt -
At where:
A0 = decay corrected specific activity, p Ci/cc A'
= specific activity attributable to fission product isotope alone (i.e.,
precursor contribution deleted), pCi/cc TABLE 15 Page 15.46
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE IS (Cont'd.) (Cont'd.)
t = the time period from reactor shutdown to sample analysis k = branching fraction for precursor decay.
X = decay constant for isotopes, sec'.
A p = decay constant for precursor, sec-.
This procedure is applied to the following isotopes from the list of selected isotopes:
Isotope Precursor Ap, secI k
Rb - 88 Kr - 88 6.8 (-5) 1.00 Xe - 133 1 - 133 9.3 (-6) 0.98 1-132 Te - 132 2.5 (-6) 1.00 Te - 129 Sb - 129 4.2 (-5) 0.83 La - 140 Ba - 140 6.3 (-7) 1.00 La - 142 Ba - 142 1.1 (-3) 1.00 Values of A' 0 for appropriate isotopes are recorded on Enclosure 6.
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.
TABLE IS Page 15.47
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
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) = A0 (
I ci/cc) x RCS Volume (STP) x 10-6 ci/
Ci where:
Ao = the specific activity of the reactor coolant sample recorded on, pci RCS Volume (STP) = the full reactor coolant system water volume corrected to standard temperature by multiplying by the factor for water density provided in.
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.
7.7.2 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.
7.7.2.2 The quantity of fission products in the sump is calculated by the following equation.
Total activity, Ci =Ao ci/cc) x Sump Water Volume x 10-6 ci/
Ci*
where:
Ao = the specific activity of the containment sump sample recorded on,
p c'/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.
(P + 14.7 492 Gas Volume (STP) = Gas Volume x I 14.7 (T
+ 460) where:
T = Containment atmosphere temperature, OF, from on Enclosure 2.
P = Containment atmosphere pressure, psig, from Enclosure 2.
- or A'0 if appropriate.
TABLE 15 Page 15.48
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
7.7.3.2 The quantity of fission products in the containment is calculated by the following equation.
Total Activity, Ci = Ao (
i/cc) x Gas Volume (STP) x 10- 6 Ci/
p Ci where:
A = The specific activity of the containment building sample recorded on,
p cl/cc.
7.7.4 The total quantity of fission products available for release to the environment is equal to the sum of the values obtained from each sample location. Record the total quantity of fission products on Enclosure 8.
7.8 Plit 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.
To correct the source inventory, the following equation is employed.
The entire 30 days power history should be employed. Enclosure 10 is provided as a worksheet.
0 Power Correction Factor =
I 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
= decay constant for isotope, se&'
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. 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 II is provided as a worksheet.
7.10 Assessment of Core Damage The conclusion on core damage is mode 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).
- or A' 0 if appropriate.
TABLE 15 Page 15.49
Southern California Edison Company Nuclear Engineering,. Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
- 3.
Quantity of the fission produce available for release to the environment expressed as a percent of source inventory (Enclosure I1).
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 made.
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 after a double-ended cold leg break, with no post-accident cooling of the core.
Use of the procedure following an accident 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 (R-1255 and R-1257). Take the average 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 12, 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 12 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 frequently as practicable.
TABLE 15 Page 15.50
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE!
Radiological Characteristics of NRC Categories of Fuel Damage Release of Characteristic NRC Category of Mechanism of Source of Characteristic Isotope Expressed as a Fuel Damage Release Release Isotopes Percent of Source Inventory I.
No Fuel Damage Halogen Spiking Gas Gap Nominal iodines Less than I Tramp Uranium and fission gases
- 2.
Initial Cladding Gas Gap Less than 10 Failure
- 3.
Intermediate Clad Burst and Gas Gap Xe 131m, Xe 133 10 to 50 Cladding Failure Gas Gap Diffusion 1 131,1 133, Kr-88 Release Kr 85m
- 4.
Major Cladding Gas Gap Greater than 50 Failure
- 5.
Initial Fuel Pellet Fuel Pellet Cs-136, Rb 88, Less than 10 Overheating Te 129, Te 132
- 6.
Intermediate Grain Boundary Fuel Pellet 10 to 50 Fuel Pellet Diffusion Overheating
- 7.
Major Fuel Pellet Diffusional Release Fuel Pellet Greater than 50 Overheating From U0 2 Grains
- 8.
Initial Fuel Pellet Fuel Pellet Less than 5 Melt
- 9.
Intermediate Fuel Escape from Molten Fuel Pellet Ba 140, La 140, 5 to 20 Pellet Melt Fuel
- 10.
Major Fuel Pellet Fuel Pellet Greater than 20 Melt TABLE 15 Page 15.51
Southern California Edison Company Nuclear Engineering, Safety & Licensing Oept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 2 Record of Plant Indications Date Time Reactor Coolant System Pressure
._psig Temperature OF Vessel Level Pressurizer Level 9%
RCS Volume 1.86 x 108 cc Containment Building Pressure psig Temperature
-OF Sump Level Ft Containment Volume 3.4 x 1010 cc Prior 30 Days Power History Power for days,then Power for days, then Power for days,then Power for days to time of shutdown Reactor Shutdown Time Date TABLE 15 Page 15.52
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 3 Record of Sample Specific Activity Sample Number:
Location:
Time of Ana!ysis:
Temperature, OF:
Pressure, PSIG:
Isotope Sample Activity, pcI/cc Kr 85m Kr 88 Xe 131m Xe 133 1 131 1 132 1 133 1 135 Cs-136 Rb 88 Te 129 Te 132 Ba 140 La 140 TABLE IS Page 15.53
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 4 DENSITY CORRECTION FACTOR FOR REACTOR COOLANT TEMPERATURE T, oF 800 700 600 500 400 300 200 100 0
(
0.25 0.50 0.75 1.00 1.25 Factor TABLE 15 Page 15.54
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 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 Achvity Isotope (Enclosure 3), oc'/cc Factor at STP, #6cI/cc Kr 85m Kr 88 Xe 131m Xe 133 1 131 1 132 1 133 1 135 Cs-136 Rb 88 Te 129 Te 132 Ba 140 La 140 TABLE 15 Page 15.55
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/M5/85 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 6 Record of Sample Decay Correction Sample Number:
Location:
Time of Reactor Shutdown:
Time of Analysis:
Elasped Time, t:
Precursor Decay Specific Activity Decay Corrected Corrected
- Constant, at STP (Enclosure 5),
Specific Activity, Specific x
JA)
(A Activity Isotope I/sec Aclc
/cc (A'A)Aci/cc Kr 85m 4.3 (-5)
Kr 88 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-I 36 6.3 (-7)
Rb 88 6.5 (-4)
Te 129 1.7 (-4)
Te 132 2.5 (-6)
Ba 140 6.3 (-7)
La 140 4.8 (-6)
TABLE 15 Page 15.56
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV.8 04/15/85 TABLE IS (Cont'd.) (Cont'd.)
ENCLOSURE 7 Record of Fission Product Release Source Identification Sample Number:
Location:
Decay Corrected Specific Activitiy Calculated Activity Ratio Activity Ratio Identified (Enclosure 6). Ac.c'/cc Isotope Ratio*
in Fuel Pellet In Gas Gap Source Kr 85m 0.1 0.02 Kr 88 0.2 0.001 Xe 133 1.0 1.0 1.0 1 131 1.0 1.0 1.0 1 132 1.4 0.01 1 133 2.0 0.5 1 135 1.8 0.17
- Noble Gas Ratio Decay Corrected Noble Gas Specific Activity Decay Corrected Xe 133 Specific Activity or
- Iodine Ratio
= Decay Corrected Iodine Isotope Specific Activity Decay Corrected 1 131 Specific Activity TABLE 15 Page 15.57
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 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 Isotope Ci Ci Number, Ci Ci Kr 85m Kr 88 Xe 131m Xe 133 I 131 1 132 1 133 1 135 Cs-136 Rb 88 Te 129 Te 132 Ba 140 La 140 TABLE 15 Page 15.58
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/IS/85 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 9 CONTAINMENT BUILDING SUMP LEVEL LEVEL FEET
+2 07
-2
-10.
0 2
810 12 VOLUME, CCx10O6 TABLE 15 Page 15.59
Southern California.Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 10 Record of Correction for tbbnconstant PoAer Level (See Enclosure 2 for Po.wer History)
Decay Power Equilibriun Corrected Constant Correction Source Source Isotope
( A) I/sec Factor, F Inventory, Ci invento ly Gas Gap Inventory Kr 85rn 4.3 (-5) 1.6 (5)
Kr 88 1.5 (-4) 9.5 (4)
Xe 131m 6.7 (-7) 2.6 (4)
Xe 133 1.5 (-6) 7.2 (6) 1131 9.9 (-7) 3.6 (6) 1132 8.4 (-5) 3.9 (3) 1133 9.3 (-6) 3.5 (6) 1135 2.9 (-5) 6.4 (5)
Fuel Pellet Inventory Kr 85n 4.3 (-5) 8.2 (6)
Kr 87 1.5 (-4) 1.5 (7)
Xe 131m 6.7 (-7) 2.4 (5)
Xe 133 1.5 (-6) 7.6 (7) 1131 9.9 (-7) 3.7 (7) 1132 8.4 (-5) 5.4 (7) 1133 9.3 (-6) 7.6 (7) 1135 2.9 (-5) 6.8 (7)
Cs-136 6.3 (-7) 1.1 (6)
Fb 88 6.5 (-4) 2.2 (7)
Te 129 1.7 (-4) 1.2 (7)
Te 132 2.5 (-6) 5.4 (7)
Ba 140 6.3 (-7) 6.5 (7)
La 140 4.8 (-6) 6.8 (7)
- 1. P
(-e-At) e-to F = I I
II T
5100 TABLE I5 Page 15.60
Southern California Edison Company Nuclear Engineering. Safety& Licensing Dept.
REV. 8 04/15/85 TABLE IS (Cont'd.) (Cont'd.)
ENCLOSURE II Record of Percent Release Total Quantity Power Corrected Available for Release Source Inventory, Isotope (Enclosure 8). Ci Ci (Enclosure 10 or I1)
Percent Gas Gap Inventory Kr 85m Kr 88 Xe 131m Xe 133 1 131 1 132 1 133 1 135 Fuel Pellet Inventory Kr 85m Kr 87 Xe 131m Xe 133 1 131 1 132 1 133 1 135 Cs-136 Rb 88 Te 129 Te 132 Ba 140 La 140 TABLE 15 Page 15.61
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 12 CONTAINMENT MONITOR DOSE RATE VS. TIME AFTER INITIAL RELEASE FOR VARIOUS PERCENTAGES OF CLAD FAILURE OR CORE MELT a
S IC?
5 8
IG ll 4
I tl 2C 22 24 tat (14 AS)
TABLE 15 Page 15.62
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.)
San Onofre Unit I Procedure For Core Damage Assessment Using Hydrogen Gas 1.0 PURPOSE This procedure is employed under post accident plant conditions to estimate the extent of fuel clad damage which may have occurred utilizing hydrogen concentrations in samples of the containment atmosphere.
The measured hydrogen is related to the amount of fuel clad oxidation.
2.0 REFERENCES
2.1 "Post Accident Hydrogen Production and Buildup at San Onofre Nuclear Generating Station Unit I,"
WCAC-9636, November, 1979.
3.0 DEFINITIONS None 4.0 PRECAUTIONS AND LIMITATIONS 4.1 The assessment of core damage obtained by using this procedure is only an estimate. This procedure is based on hydrogen data. Other plant indications may be available which can improve upon estimation of core damage.
These include radiological sample characteristics and containment radiation monitors. Whenever possible, these additional indicators should be factored into the assessment.
4.2 This procedure relies upon hydrogen samples taken from the containment atmosphere. Those samples may contain a mixture of hydrogen generated within the core by clad oxidation and also hydrogen from radiolytic dissociation of water and oxidation of aluminum and zinc in the containment. The estimate of clod damage is influenced by the amount of hydrogen generated by ex-core sources and by the ability to identify plant conditions conducive to such hydrogen generation.
Therefore, a hydrogen measurement is not a unique indicator of the amount of core clad oxidation.
4.3 Depending on the accident scenario, a given total amount of hydrogen produced by oxidation of fuel clod can represent varying local amounts and distributions of clad damage. Actual clad damage could be greater than that estimated by this procedure, depending on the accident scenario.
5.0 INITIAL PLANT CONDITIONS AND SYMPTOMS This procedure is to be employed for analysis of hydrogen sample data when it is determined that a plant accident with the potential for core damage has occurred. The following is a list of plant symptons to assist in this determination.
This list is not a complete representation of all events which may indicate potential core damage.
However, the existence of one or more of these events or conditions indicates 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.
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.
TABLE IS Page 15.63
0 0
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
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 or superheated.
6.0 PREREQUISITES The capability to obtain samples of containment atmosphere and to analyze these to determine the concentration of hydrogen.
7.0 PROCEDURE 7.1 Record the Following Plant Indicators Core damage can occur following reactor trip only when the coolant level within the reactor vessel drops below the top of the active fuel. Two instrument records are available from which an estimate of the core uncovery and recovery times might be made. The instruments are:
Core Exit Thermocouple Temperature Subcooling Meter Reading Record data from these instruments according to the instructions of Enclosure I.
7.2 Obtain a gas sample from the containment atmosphere and analyze it for hydrogen concentration in volume percent.
Record the results on Enclosure I. Using Enclosure 2, estimate the percentage of cladding that has been oxidized by clod-water reaction and record result on Enclosure 1.
TABLE 15 Page 15.64
Southern California Edison Company Nuclear Enginieering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE I CORE UNCOVERY CONDITIONS Step 7.1 Time period of core uncovery. Complete the following table using recorded instrument data.
Estimated Estimated Instrument Core Uncovery Time Core Recovery Time Core Exit Thermocouple Start of Continuous Rise or Rapid Temperature Temperature Exceed 6600F.
Drop to Saturation.
Time Time Temperature Temperature Subcooling Meter Start of Superheat.
Return to Saturation or Saturation Margin Time Subcooling.
Time Step 7.2 Clad condition evaluation Volume % hydrogen v/o Percent clod oxidation TABLE IS Page 15.65
Southern California Edison Company Nuclear Engineering, Safety & Licensing Dept.
REV. 8 04/15/85 TABLE 15 (Cont'd.) (Cont'd.)
ENCLOSURE 2 CLAD-WATER REACTION VS HYDROGEN CONCENTRATION 100 0
0 20 4
101.4 1
Aj49 FANULAR STATIQ !
VOLUME PERCENT HYDROGEN TABLE 15 Page 15.66