ML20024C094
| ML20024C094 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 06/30/1983 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20024C091 | List: |
| References | |
| STATION-9918.1, NUDOCS 8307120248 | |
| Download: ML20024C094 (47) | |
Text
STETION 1043.1F1 STATION 9918.1
. REVISION 1 ORIGINAL MIDLAND PLANT i
CHP SUPERINTENDENT l
ESTIMATION OF EXTENT OF CORE DAMAGE i
Table of Effective Pages Page Revision Page Revision Page Revision Page Revision i
Original 13 Original El.1 Original E6.2 Original 1
Original 14 Original E2.1 Original E6.3 Original 2
Original 15 Original E3.1 Original E6.4 Original 3
Original 16 Original E3.2 Original E6.5 Original 4
Original 17 Original E3.3 Original E6.6 Original 5
Original 18 Original E3.4 Original E7.1 Original 6
Original 19 Original E3.4a Original 7
Original 20 Original E3.5 Original 8
Original 21 Original E3.6 Original 9
Original 22 Original E3.7 Original 10 Original 23 Original E4.1 Original
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11 Original
_ 24 Original E5.1 n < px1 y
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12 Original 25 Origimi.N. D m
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Prepared and Submitted J
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3 I driginator
'Date Heatf of Dept Date 3.
THIS IS/M A Q-LISTED PROCEDURE
[bem, d
(/C#M3 Chair:nda PRC Dste '
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PRC Recommends Approval 5.
Reviewed l>
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6.
Approved l
l Nc3 Safety Evaluation Dated 9-/ f-83 l
N Plant Manager Date st1182-4113i165-12
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4
. STATION 1043.1F2 2
REVISION 2 i
Identification Number STATION 9918.1 Title ESTIMATION OF Er1T.NT OF CORE DAMAGE Cognizant Department Head CHP Superintendent i
The following paragraphs have "(later)" within,them. Those identified as
~" Resolve Prior to Use" shall be so resolved. Remaining items shall be incorporated in the first revision following their resolutio.
M (M #J General Manager Date Paragrag Resolve Prior to Use Paragraph Resolve Prior to Use I
3.10 Yes 3.11 Yes 8.2.la Yes I,
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8.2.lb Yes i
8.2.le Yes i
-8.2.lf Yes l
3.2.1g l
Yes I
8.2.3 Yes Enc 3 Yes l
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i' st1182-4113jl65-12
A STATION 9918.1 ORIGINAL Page i TABLE OF CONTENTS
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TOPIC PAGE 1.
PURPOSE.
I 2.
APPLICABILITY.
1 3.
REFERENCES.
1 4.
GENERAL INFORMATION.
3 5.
PREREQUISITES / INITIAL CONDITIONS.
3 6.
PRECAUTIONS.
3 7.
LIMITATIONS AND ACTIONS.
5 8.
PROCEDURE.
5 8.1 Method A-Estimating of Extent of Core Damage From Containment Area Radiation Monitors.
4 8.2 Method B-Estimating of Extent of Core Damage Utilizing Samples Obtained from PAS.
6 9.
CHECK 0FF LISTS.
24 10.
ENCLOSURES,
. 24 11.
DISTRIBUTION.
. 25.
ENCLOSURES i
ENCLOSURE TITLE I
Data Sheet for Determining Extent of Core Damage (Method A) 2-Correct Exposure Rate vs Elapsed Time 3
Data Sheet for Determining the Extent of Core Damage (Method B) 4 Nuclide Parameters 5
Core Inventory Correction Factors 6
Core Exit Fluid Temperature for Inadequate Core Cooling 7
Flowchart Method B st1182-4113k165-12
STATION 9918.1 ORIGINAL Page 1 ESTIMATION OF EXTENT OF CORE DAMAGE 1.
PURPOSE To provide methods of estimating the extent of core damage during and af ter an accident utilizing containment area radiation monitors or samples obtained from the Post-Accident Sampling System (PASS).
2.
APPLICABILITY o
2.1 Method A shall be used by the Technical Support Center Staff in the assessment of the amount of core damage during the evolution of an accident or after an accident has occurred. This method may be used without obtaining samples from PASS.
2.2 Method B is to be used under the direction of the Radiological Assessment Coordinator upon completion of the post-accident sample analysis. This method will produce a more accurate estimate of the extent of core damage and shall be the preferred method.
2.3 The Site Emergency Planning Coordinator, under the direction of the CEP Superintendent, is responsible for maintenance and review of this procedure.
2.4 This procedure is only applicable after issuance of an operating license and when the Site Emergency Plan has been implemented.
3.
REFERENCES 3.1 Drawing J-3115, Instrument Location Drawing Area 1 Plan of Elevation 659-0 J
st1182-4113a165-12
o STATION 9918.1 CRIGINAL Page 2 4
3.2 Drawing J-3116, Instrument Location Drawing Area 1 Plan of Elevation 659-0
- 3 Drawing J-3127, Instrument Location Drawing Reactor Building Area 2 Plan of Elevation 659-0 3.4 Drawing J-3153, Instrument Location Drawing Reactor Building Area 4 Plan of Elevation 659-0 3.5 Drawing J-3143, Instrument Location Drawing Area 3, Plan of Elevation ~659-0 3.6 Drawing J-3144, Instrument Location Drawing Area 3 Plan of Elevation 659-0 3.7 FSAR, Table 15A-2 3.8 NRC Regulatory Guide 1.4 3.9 STATION 1502.1, Records Collection, Storage and Maintenance 3.10 STATION 9915.2, Post-Accident Sampling, Reactor Coolant and Containment Sump (Later) 3.11 STATION 9915.3, Post-Accident sampling, Containment Atmosphere (Later) 5.12 STATION 9915.5, Post-Accident Samples Analyses 3.13 FSAR Chapter 11.1, Source Terms st1182-4113a165-12
r-STATION 9918.1
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ORIGINAL Page 3 4.
GENERAL INFORMATION 4.1 The core and fuel gap source terms used in Method B are based on FSAR Chapter 11.1 calculations and assumptions. The core nuclide activity values are based on core full power operation at 2552 MWt during a 310 effective full power day equilibrium cycle.
4.2 A flow chart detailing Method B is provided in Enclosure 7.
5.
PREREQUISITES / INITIAL CONDITIONS 5.1 For Method A, at least one of the two monitors for the applicable unit listed in Section 8.1.1 is required to be operational and on-scale.
5.2 For' Method B, the ability to analyze post-accident samples for various radionuclides is required.
6.
PRECAUTIONS 1
6.1 Method A 6.1.1 Depending on the time in core life and the operating history, this procedure may give slightly nonconservative results.
(This procedu're is based on equilibrium core activities from FSAR Table 15A-2, operation at 100% power for the duration of the fuel cycle, anc NRC Regulatory Guide 1.4 assumptions for core releases.)
6.1.2 Results should be compared with the results from Method 3 of this procedure, if available.
st1182-4113a165-12
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STATION 9918.1 ORIGINAL Page 4 6.1.3 This method is dependent upon the reactor coolant boundary being violated. Post-accident sample results are needed to estimate core damage if primary coo,lant laakage has not occurred.
t 6.1.4 This method assumes that the entire inventory of gaseous fission products are released to the primary coolant. This method assumes a LOCA has occurred and cooling is in the recirculation mode.
If, in fact, the accident is a small break LOCA that is quickly contained, the amount of fission products released to containment will be less than 100%;
therefore, this method will underestimate the extent of core damage.
6.2
. Method B 6.2.1 Due to the short half-lives of Kr-35m (4.48 h) and Sr-92 (1.71 h), these nuclide indicators should only be measured and used in this method during the first twelve hours after reactor shutdown.
6.2.2 This procedure provides only an estimate of the extent of fuel rod failure, fuel overheating, and fuel melting.
Therefore, results from this method should not be interpreted as being absolute values for core damage but as a "best estimate" based on information available immediately following an accident.
st1182-4113s165-12
STATION 9918.1-ORIGINAL Page_5 7.
LIMITATIONS AND ACTIONS 7.1 None.
8.
PROCEDURE ~
8.1 Method-A
Estimating the extent of core damage from containment area radiation monitors.
8.1.1 For the applicable unit, determine the pre-accident and post-accident exposure rates (use recorders in control room if available or histogram information from Digital Radiation Monitoring System) from both of the following containment area radiation monitors:
Unit Detector Number 1
1RE-8418 1
IRE-8440 2
2RE-8419 2
2RE-8441 Record the pre-accident and post-accident exposure rates on 8.1.2 Determine the elapsed time from the time of shutdown to the time at which the post-accident exposure rate is measured.
Enter the elapsed time on Enclosure 1.
I 8.1.3 Determine the corrected exposure rate by subtracting the pre-accident exposure rate from the post-accident exposure L
O s 1182-4113a165-12
STATION 9918.1 ORIGINAL Page 6 s
rate for the same monitor. Enter this as the corrected exposure rate on Enclosure 1.
8.1.4 Using Enclosure 2, use the elapsed time and corrected I
exposure rate to determine the approximate percent of fuel failure.
If the fuel failure falls below 1%, multiply the corrected exposure rate by 10 and redetermine the percent of fuel failure as above. Divide this percent by 10.
Eater the results on Enclosure 1.
8.1.5 Initial and date in the provided space on Enclosure 1.
3.1.6 After recording the results, submit Enclosure I for review and signature by the Health Physicist and the Site Emergency Director.
8.2 Method B - Estimating the extent of core damage utilizing samples obtained from PASS.
8.2.1 Determine and record the following plant parameters from j
data recorded at reactor shutdown and at subsequent hourly intervals up until the time of post-accident sampling.
Only record the value at the time of post-accident sampling if so stated.
8.2.1.a For the applicable unit, view the Safety parameter Display System (SPDS) display (Later),
" Core Exit Fluid Temperature for Inadequate Core Cooling" diagram at the time of post-accident st1182-4112a165-12
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STATION 9918.1 ORIGINAL Page 7 sampling or as soon as possible thereafter.
Determine the operation ' region ~ and enter the region number in Enclosure 3 (Page 1).
8.2.1.b For the applicable unit, record in Enclosure 3 (Page 2) the core exit thermocouple average temperature values from the SPDS computer display (Later).
In the case where the SPDS display is not available, obtain the current temperature reading from the Post-Accident Monitoring (PAM) panel 1/2C-31 and enter in Enclosure 3 (Page 2).
Enter source of temperature value in Enclosure 3 (Page 2).
8.2.'.c For the applicable unit, record in Enclosure 3 (Page 2) the reactor coolant system pressure values from the plant computer Post-Trip display (Later). In the case where the computer display i
l is not available, obtain the current pressure l
reading from the PAM panel display 1/2C-31 and l
enter in Enclosure 3 (Page 2).
Enter source of I
pressure value in Enclosure 3 (Page 2).
l l
8.2.1.d Using Enclosure 6, Figure 4, determine the i
operation region for the reactor core at each time increment using the average core exit i
l thermocouple temperature and RCS pressure values e
st1182-4113a165-12 l
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STATION 9918.1
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ORIGINAL Page 8 J
obtained in Steps 8.2 1.b and 8.2.1.c and enter in Enclosure 3 (Page 2).
8.2.1.e Using the result from Steps 8.2.1.a and 8.2.1.d determine the most severe operating condition attained between the time of reactor shutdown and
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post-accident sampling. The most severe
-operating condition corresponds to the highest numbered core operation region determined in
' Steps 8.2.1.a and 8.2.1.d.
Enter result in (Page 1).
8.2.1.f For the applicable unit, determine the contginment hydrogen volume percent frca the plant computer display (Later) and enter in (Page 3).
In the case where the computer is unavailable, obtain the volume percent value from the PAM panel display 1/2C-31 and enter in Enclosure 3 (Page 3).
Enter source of value in Enclosure 3.(Page 3).
8.2.1.g For the applicable unit, determine the i
containment radioactivity levels from the high range monitors (IRE-8418, 1RE-8440 or 1RE-8419, 2RE-8441) from the RMS computer display (Later) and enter in Enclosure 3 (Page 3).
In the case where the computer is uncvailable, obtain the radiation level frem Control Panel OC403 display st1132-41134165-12 1
t-
o e
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STATION 9918.1 ORIGINAL Page 9 and enter in Enclosure 3 (Page 3).
Enter source of value' in Enclosure 3 (Page 3).
8.2.2 Perform Sampling and Analysis 8.2.2.a For the applicable unit, perform post-accident sampling as described in Reference 3.10 and 3.11, (Post-Accident Sampling, Reactor Coolant and
(
Containment Sump) and 3.11 (Post-Accident Sampling, Containment Atmosphere).
8.2.2.b Analyze reactor coolant samples as described in Reference 3.12, Post-Accident Samples Analyses.
The measured nuclide concentrations will be adjusted to take into account radioactive decay from the time of reactor shutdown. Enclosure 4 is provided to accomplish this.
8.2.2.b.1 Obtain the measured I-131 concentration (pCi/g) in the reactor coolant. Correct to time of shutdown using Enclosure 4 and enter the measured and corrected values in (Page 4).
8.2.2.b.2 Obtain the measured Cs-137 concentration (pCi/g) in the reactor coolant. Correct to time of shutdo'.n using Enclosure 4 and enter the st1182-4113a165-12
o STAT 20N 9918.1 ORIGINAL Page 10 measured and corrected values in '(Page 4).
8.2.2.b.3 Obtain the measured La-140 concentration (pCi/g) in the reactor
(
coolant. Correct to time of shutdown using Enclosure 4 and enter the measured corrected values in (Page 4).
8.2.2.b.4 Obtain the measured St-92 concentration (pCi/g) in the reactor.
coolant. Correct to time of shutdown using Enclosure 4 and enter the measured and corrected values in (Page 4).
8.2.2.b.5 Obtain the =essured Sr-91 concentration (pCi/g) in the reactor coolant. Correct to time of shutdown using Enclosure 4 and enter the measured and corrected values in (Page 4).
8.2.2.b.6 Obtain the total concentration in the reactor coolant. This will be the sum of all radionuclides present st11S2-4113a165-12
O STATION 9918.1 ORIGINAL Page 11 corrected-to time of shutdown. Enter this value in Enclosure 3 (Page 4).
8.2.2.b.7 Compare the La-140, Sr-91, and Sr-92 e
corrected concentrations determined in Steps 8.2.2.b.3, 8.2.2.b.4 and 8.2.2.b.5 with the 50% core gap release values shown in Step 8.2.3.d.2.
If any one of the corrected nuclide concentrations is equal to or greater than the 50% core gap releases values, enter "yes" in (Page 4).
+
8.2.2.b.8 Obtain the measured hydrogen gas concentration (cc/kg) in the reactor coolant. Enter measured value in (Page 4a).
l 8.2.2.c Analyze containment sump sample as described in Reference 3.12, Post-Accident Sample Analyses.
8.2.2.c.1 Obtain the measured La-140 concentration (pci/g) in the containment sump sample. Correct to time of shutdown using Enclosure 4 I
and enter the measured and corrected values in Enclosure 3 (Page 5).
l l
st1132-4113a165-12
1 STATION 9918.1
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ORIGINAL Page 12
_s i
8.2.2.c.2 Obtain the measured Sr-92 concentration (pCi/g) in the containment sump sample. Correct to time of shutdown using Enclosure 4 and enter the measured and corrected values in Enclosure 2 (Page 5).
8.2.2.c.3 Obtain the measured Sr-91 concentration (pci/g) in the e
containment sump sample. Correct to time of shutdown using Enclosure 4 and enter the measured and corrected values in Enclosure 3 (Page 5).
8.2*.2.d Analyze containment atmospbere sample as described in Reference 3.12, Post-Accident Samples Analyses.
8.2.2.d.1 Obtain the measured Xe-133 concentration (pCi/cc) in the containment atmosphere. Correct to time of shutdown using Enclosure 4 I
and enter the measured and corrected values in Enclosure 3 (Page 5).
8.2.2.d.2 Obtain the measured Kr-85m concentration (pCi/ce) in the containment atmosphere. Correct to s t1182-4113a165-12
8 O
STATION 9918.1 ORIGINAL Page 13 time of shutdown using Enclosure 4 and enter the measured and corrected values in Enclosure 3 (Page 5).
8.2.3 Determine Core Condition The following section contains a description of each of the four core conditions: normal operation, observable macroscopic clad damage, severe fuel overheating, and fuel e
melting. The following steps will be used to identify the core condition based on plant information previously measured and recorded in Steps 8.2.1 and 8.2.2.
Make the appropriate entries in Enclosure 3 (Page 6) using the information in the fol1 wing Steps 8.2.3.a.2, 8.2.3.b.2, 9
8.2.3.c.2, and 8.2.3.d.2.
Identify the core conditions by the existence of at least two out of three indications present for Conditions I, III, and IV.
Condition II will be identified only if all three indications are present.
Make the appropriate entry at the bottom of Enclosure 3 (Page 6).
If more than one is identified, enter the most severe condition.
Determine how many effective full power days of operation i
have occurred in the current cycle at the time of reactor shutdown from the plan' computer display (Later). Enter effective full power days in Enclosure 5 (Page 1).
Use - (Page 1) with interpolatica techniques to obtain core inventory correction factors and enter the st1182-41132165-12 l
I I
STATION 9918.1 ORIGINAL lege 14 e
+g factors in Enclosure 5 (Page 1).
Depending on the core condition identified at the bottom of Enclosure 3 (Page 6),
estimate the percent failed fuel, percent fuel overheated, or percent fuel melted using the appropriate procedure in i
Steps 8.2.3.a.3, S.2.3.b.3, 8.2.3.c.3, and 3.2.3.d.3.
8.2.3.a Condition I - Normal Operation 8.2.3.a.1 Conditions of Normal Operation:
Normal reactor operation at any power, or shutdown with no unusual conditions prior to shutdown.
Adequate core cooling has been
+
maintained. All core exit thermocouple temperatures indicate subcooled reactor coolant (operation in Region 1 of Enclosure 6, Figure 4).
3.2.3.a.2 Indications of Normal Operation:
(a) Total concentration in reactor coolant is less than 200 pCi/g (17. failed fuel).
(b) Reactor coolant and containment hydrogen levels at or below st1182-4113a165-12 L
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f STATION 9918.1 ORIGINAL Page 15 normal operating levels (50 cc/kg, ~ 0 Vol %).
(c) All core exit thermocouple temperatures along with reactor coolant system pressure indicate operation in subcooled region of, Figure 1 (Region 1).
8.2.3.a.3 Estimation of % Failed Fuel Divide the I-131 concentration determined in Step 8.2.2.b.1 by the appropriate core inventory correction factor in Enclosure 5 (Page 1).
Enter the adjusted value in (Page 7).
Make an estimation of % failed fuel using, Figure 2.
Enter in (Page 7).
8.2.3.b Condition II - Observable Macroscopic Clad Damage 8.2.3.b.1 Conditions of Observable Macroscopic Clad Damage:
Normal reactor operation at any power or shutdown where some mechanical st1182-4113a165-12
STATION 9918.1 ORIGINAL Page 16
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clad failure is indicated. The coee has adequate cooling and no significant fuel over temperature is observed. All core exit thermocouple temperatures indicate operation in subcooled Region 1 or Region 2 (superheated steam but not severe enough to cause cladding damage) as shown on Enclosure 6, Figure 1.
8.2.3.a.2 Indications of Observable Macroscopic Clad Damage (a) Total concentration in reactor coolant is greater than 200 pCi/cc (1% failed fuel level).
(b) Reactor coolant and containment hydrogen levels at or below normal operating levels (50 cc/kg, ~ 0 Vol %).
(c) Core exit thermoccuple temperatures along with reactor coolant system pressure indicate operation in Region 1 or 2 as shown on Enclosure 6, Figure 1.
st1182-4113a165-12
3 STATION 9918.1 ORIGINAL.
Page 17 8.2.3.b.3 Estimation of ?. Failed Fuel Divide the I-131 concentration determined in Step 8.2.2.b.1 by the appropriate core inventory correction
~
factor in Enclosure 5 (Page 1).
~
Enter the adjusted value in
, (Page 7). Make an estimation of % failed fuel using, Figure 3.
Enter in
- '~~ ~~ ~~ (Page 7).
8.2.3.c Condition III - Severe Fuel Overheating S.2.3.c.1 Conditions of Severe Fuel Overheating:
Abnormal shutdown conditions where it is suspected that the fuel has been-at least partially uncovered for~a period of time greater than a few minutes. Voiding in the core is detected by high core exit thermocouple temperatures and loss of reactor coolant saturation margin l -
indicated by operation in Region 3 of, Graph 1.
Fuel clad oxidation is detected by excess st1132-4113a165-12 l
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I s
STATION 9918.1 s
ORIGINAL Page 18 l
hydrogen in the contair.:;:ent atmosphere or in the reactor coolant; no fuel melting is suspected.
1 8.2.3.c.2' Indications.cf Severa Fuel I
Overheating:
s (a) High range containment,.ionitors 1RE-8418,1RE-8440,2E5-8419,or 2RE-8441 scow post-accident radiation levels greater than
'1000 R/hr during the first 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after shutdown.
(b) Reactor coolant hydrogen concentration greater than 50 cc/kg or containment hydrogen levels indicative of greater than 10% clad-water interaction
(.58 vol % H2 pst-LOCA steam environment or 2.47 vol % H d#7 2
containment).
(c) Core exit thermocouple temperatures along with reactor coolant system pressure indicate operation in Region 3 of e
st1182-4113a165-12
E
]
c STATION 9918.1 ORIGINAL Page 19
., Figure 1 for extended periods.
]
8.2.3.c.3 Estimation of % Core Overheated 1
Fuel overheating can occur after two different LOCA scenarios; isolatable or nonisolatable. An isolatable LOCA is when the source of primary coolant loss has been isolated early in the accident with minimal coolant spillage into the reactor building sump. A nonisolatable LOCA is when the source of primary coolant leakage has not been isolated and continual loss of primary inventory to the reactor building sump occurs. Core cooling would be reestablished in the i
piggy-back mode of operation:
coolant injection by the HPI System and recirculation through the reactor building stuep and DHR heat i-ex: hangers. Determine the % of core overheated assuming both isolatable i
and conisolatable LOCA using the following:
st1182-4113a165-12
STATION 9918.1
' ORIGINAL Page 20 (a)
Isolatable LOCA Divide I-131 and Cs-137 concentration determined in
^
Steps 8.2.2.b.1 and 8.2.2.b.2 by the appropriate core inventory correction factor in Enclosure 5 (Page 1).
Enter both adjusted values in Enclosure 1 (Page 7).
Make an estimation of %. core overheated, using Enclosure 6, s
Figure 4 for both isotopes.
Enter both values in Enclosure 3 (Page 7).
Average both values and enter in Enclosure 3 (Page 7).
(b) Nonisolatable LOCA Divide Xe-133 and Kr-85m concentration determined in Steps 8.2.2.d.1 and 8.2.2.d.2 by the appropriate core inventory correction factor in Enclosure 5 t.'Page 1). Enter both adjusted values in Enclosure 3 (Page 7).
Make an estimation of % core overheated, using Enclosure 6, st1182-4113a165-12
STATION 9918.1 ORIGINAL Page 21 Figure 5 for both isotopes.
Enter both values in Enclosure 3 (Page 7).
Average both values and enter in Enclosure 3 (Page 7).
8.2.3.c.4 Percent Core Overheated Determine which ?. core overheated l
value is greater from the two values determined for the isolatable and nonisolatable LOCA cases. Enter the greater of the two values in (Page 7).
8.2.3.d Condition IV - Fuel Melting i
l l
8.2.3.d.1 Conditions of Fuel Melting Severe accident where there has been a loss of shutdown cooling and the r-l core is uncovered for a long period t
of time. Core exit thermocouple l
l-temperatures indicate operation in Region 4 of Enclosure 6, Graph I for long periods of time. Fuel melting I
t.
is suspected (ie, fuel temperature i
j-exceeds 5,000*F).
t
(
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STATION 9918.1
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ORIGINAL Page 22 8.2.3.d.2 Indications of Fuel Melting:
(a) High range containment Monitors 1RE-8418, 1RE-8440, 2RE-3419, or 2RE-8441 show post-accident
^
radiation levels greater than 10,000 R/hr during the first 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after an accident.
Tb) Core exit thermocouple temperatures aiong with Reactor Coolant System pressure readings indicate operation in Region 4 of Enclosure 6, Graph I for extended periods.
(c) Certain core fission products are detectable in the reactor coolant or containment sump samples in excess of normal operating concentrations. This indicates nuclide releases from the fuel pellet. These nuclides include La-140, Sr-92, and Sr-91.
st1182-4113a165-12 l
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a.
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STATION 9918.1 ORIGINAL Page 23 Normal RCS Concentration (pCi/g)
(Based on 1% Failed Fuel)
~3 La-140 2.45 x 10
-3 Sr-92 1.12 x 10
-3 Sr-91 5.95 x 10 Accident RCS Concentration (pCi/g)
(Based on 50% Core Gap Release La-140 5.20 x 10+0
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Sr-92 6.45 x 10
~I Sr-91 4.54 x 10 8.2.3.d.3 Estimation of % Core Melted Divide Xe-133 and Kr-85m concentrations determined in Steps 8.2.2.d.1 and 8.2.2.d.2 by the appropriate core inventory correction factor in Enclosure 5 (Page 1).
Enter both adjusted values in (Page 7).
Make an estimation of % core melted using, Figure 6, for both isotopes. Enter both values in (Page 7). Average both st1182-4113a165-12
STATION 9918.1 ORIGINAL Page 24 values and entar in Enclosure 3 (Page 7).
.8.2,4 After recording results, submit Enclosure 3 for approval by the Site Emergency Director.
8.3 All records shall be m3intained in accordance with Reference 3.9, (STATION 1502.1, Records Collection, Storage and Maintenance.)
9.
CEECKOFF LISTS 9.-1 No ne- - - -
10.
ENCLOSURES 10.1 Data Sheet for Determining Extent of Core Damage.
10.2 Corrected Exposure Rate vs Elapsed Time.
10.3 Data Sheet for Determining the Extent of Core Damage (pp 1-7).
10.4 Nuclide Parameters (p I).
10.5 Core Inventory Correction Factors (p 1).
10.6 Graphs.
10.6.1 Core Exit Fluid Temperature for Inadequate Core Cooling (p 1).
10.6.2 Case I - Normal Operation - Expected I-131 Activity Concentration in Reactor Coolant vs Percent Failed Fuel (p 2).
stI182-4113a165-12
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STATION 9918.1 ORIGINAL Page 15 10.6.3 Case II - Observable Macroscopic Clad Damage - Expected I-131 Concentration in Reactor Coolant vs Percent Failed Fuel (p 3).
10.6.4 Case III - Severe Fuel Overheating - Expected Nuclide Concentration in Reactor Coolant vs Percent Core Overheated (p 4).
10.6.5 Case III - Severe Fuel Overheating - Izpected Nuclide Concentration in Containment Atmosphere vs Percent Core Overheated (p 5).
10.4.6 Case IV - Fuel Melting - Expected Nuclide Containment Concentration vs Percent Core Melted (p 6).
10.7 Flowchart (p 1).
11 DISTRIBUTION A.3 A.5 G1 e
st1132-4113a165-12
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, STATION 9918.1F1 ENCLOSURE 1
~
ORIGINAL STATION 9918.1 ORIGINAL Page 1 DATA SEET FOR DETERMINING EXTENT OF CORE DAMAGE (Method A)
PRE-ACCIDENT POST-ACCIDENT CORRECTED PERCENT ELAPSED MONITOR EXPOSURE EXPOSURE EXPOSURE FUEL CALCULATED
' TIME (HRS)
NUMBER RATE (R/hr)
RATE (R/hr) RATE (R/hr) FAILURE (*.)
BY/DATE 1RE-8418 1RE-8440
, 2RE-8419 2Pi-8441 1RE-8418 1RE-8440 2RE-8419 2RE-8441 Reviewed By:
Health Physicist Date Site Emergency Director Date 6 bI. d i! M $ J M 1 h d } 300 -
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r STATION 9918.1F2-1 ENCLOSURE 3 ORIGINAL STATION 9918.1 ORIGINAL Page 1 DATA SHEET FOR DETERMINING THE EXTENT OF CORE DAMAGE (Method B)
Step 8.2.la SPDS Inadequate Core Cooling Diagram (Computer Display (Later)
Time Ooerating Region RC Subcooled - Region 1 RC Superheated - Region 2 RC Superheated - Region 3 RC Superheated - Region 4 a
Step 3.2.le Final Result - Core Ooerating Condition RC Subcooled a Region 1 RC Superheated - Region 2 RC Superheated - Region 3 RC Superheated - Region 4 Performed by:
st1182-4113d165-12 Date:
STATION 9918.1F2-2 ENCLOSURE 3
~
ORIGINAL.
STATION 9918.1 ORIGINAL Page 2 Step 8.2.Ib Core Exit Thermocouple Displav 1 - RC Subcooled Step 8.2.1c (Computer Display (Later))
2 - RC Superheated Step 8.2.1d (PAM Panel Display (1/2C31) 3 - RC Superheated 4 - RC Superheated Source RCS Source Operating Time Avg Temp ('T) of Value Pressure (psia) of Value Region Performed by:
st1182-4113d165 Date:
r.
~*
STATION 9918.1F2-3 ENCLOSURE 3 ORIGINAL
-STATION 9918.1 CRIGINAL Page 3 Step 8.2.lf Containment Hydrogen Monitor (Computer Display (Later))
Time Volume %
Source of Value Step 8.2.lg High Range Containment Radiation Monitor (RMS Display (Later))
OC403 Display (Later))
Time Radiation Level (R/hr)
Source of Value' 1
Performed by:
st1182-4113d165-12 Date:
STATION 9918.1F2-4 ENCLOSURE 3
-ORIGINAL STATION 9918.1 ORIGINAL Page 4 Step 8.2.2b Reactor Coolant Sample Time of
-Sample Date I-131 Concentration (pci/g)
Cs-137 Concentration (pCi/g)
Measured / Corrected Measured / Corrected
~
i Concentration (pCi/g)
Time of Measured / Corrected Concentration >
Sample Date La-140 Sr-92 Sr-91 50% Gap Release?
l l
l l
Time of Sample Date Total Concentration (pCi/g) i 4
4 t
Performed by:
st1132-4113d165-12 Date:
' STATION 9918.1F2-5 ENCLOSURE 3 ORIGINAL STATION 9918.1 j
ORIGINAL Page 4a Step 8.2.2b Reactor Coolant Samole Time of Sample Date
~ Hydrogen Concentration (ce/kg)
+
1 a
t I
?
a f
f l
9' Performed by:
st1182-4113d165-12.
Date:
.s STATION 9918.1F2-6 ENCLOSURE 3 ORIGINAL STATION 9918.1 ORIGINAL Page 5 Step S.2.2c Containment Sump Sample La-140 Sr-92 Sr-91 Time of Concentration Concentration Concentration Sample Date (pCi/g)
(pCi/g)
(pCi/g)
Measured / Corrected Measured /Cerrected Measured / Corrected Step 8.2.2d Containment Atmosobere Samole Time of Samole Date Xe-133 Concentration (pCi/ce)
Kr-85m Concentration (pCi/ce)
Measured / Corrected Measured / Corrected i~
t l
l l
l l
l l
Performed by:
t st1182-4113d165-12 Date:
l I
_ ~ _..
I' STATION 9918.1F2-7 ENCLOSURE 3 ORIGINAL STATION 9918.1 ORIGINAL Page 6 CORE CONDITION INDICATION CHECKLIST (Check box if indication is met as described in procedure)
(Condition identified by existence of at least 2 out of 3 indications)
Step 8.2.3a2 Normal Operation - Condition I Indication 1.
Total concentration in reactor coolant < 200 pCi/g (Enclosure 3, Page 4)
Indication 2.
H levels < 50 cc/kg, - 0 Vol % (Enclosure 3, Page 3, 4a) 2 Indication 3.
Core T/C show subcooled RC (Region 1) (Enclosure 3, P2ge 1)
Step 8.2.3b2 Observable Macroscopic Clad Damage - Condition II Indication 1.
Total concentration in reactor coolant > 200 pCi/g (Enclosure 3, Page 4)
Indication 2.
H levels < 50 cc/kg, ~ 0 Vol % (Enclosure 3, Page 3, 4a) 2 Indication 3.
Core T/C show subcooled RC (Region 1) (Enclosure 3, Page 1) or superheated RC (Region 2)
Step 8.2.3c2 Severe Fuel Overheating - Condition III Indication 1.
High range containment monitor > 103 R/hr and < 104 R/hr (Enclosure 3, Page 3)
Indication 2.
H levels > 50 cc/kg or >.58 Vol % steam environment 2(or 2.47 Vol % dry environment) (Enclosure 3, Page 3, 4a)
Indication 3.
Core T/C show superheated RC (Region 3) (Enclosure 3, Page 1)
Step 8.2.3d2 Fuel Melting - Condition IV Indication 1.
High range containment monitor > 104 R/hr (Enclosure 3, Page 3)
Indication 2.
Core T/C show superheated RC (Region 4) (Enclosure 3, Page 1)
Indication 3.
La-140, Sr-92 or Sr-91 RCS concentrations greater than 50% core gap release values (Enclosure 3, Pahe 4)
Core Condition I. Normal Operation II. Observable Macroscopic Clad Damage III. Severe Fuel Overheating IV. Fuel Melting Performed by:
st1182-4113d165-12 Date:
g-
. s STATION 9918.1F2-8 ENCLOSURE 3 ORIGINAL STATION 9918.1 ORIGINAL Page 7 FINAL RESULTS Step 8.2.3a3 Normal Operation - Case I Adjusted I-131' Activity (pCi/g)
% Failed Fuel (From Figure 2)
Step 8.2.3b3 Macroscopic Clad Damage - Case II Adjusted I-131 Activity (pCi/g)
% Failed Fuel (From Figure 3)
Step 8.2.3c3 Severe Fuel Overheating - Case III Isolatable LOCA Adjusted I-131 Activity (pCi/g)
Adjusted Cs-137 Activity (pci/g)
% Core Overheated (From Figure 4)
/
% Core Overheated (Average)
Nonisolacable LOCA Adjusted Xe-133 Activity (pCi/ce)
Adjusted Kr-85m Activity (pCi/cc)
% Core Overheated (From Figure 5)
/
% Core Overheated (Average)
Step 8.2.3c4 % Core Overheated Step 8.2.3d3 Fuel Melting - Case IV Adjusted Xe-133 Activity (pCi/ce)
~
Adjusted Kr-85m Activity (pCi/cc)
% Core Melted (From Figure 6)
/
% Core Melted (Average)
Performed by:
Date:
Reviewed by:
Engineering Coordinator st1182-4113d165-12 Approved by:
(SED)
.s ENCLOSURE 4
~
STATION 9918.1 ORIGINAL Page 1 NUCLIDE PAR E TERS
~
Half Life (Hour)
Decay Constant (Hour )
-3 I-131 192.98 3.592 x 10 5
~0 Cs-137 2.64 x 10 2.626 x 10
~
-2 La-140 40.23 1.723 x 10
-2 Sr-91 9.52 7.281 x 10
~1 Sr-92 2.71 2.558 x 20
-3 Xe-133 126 5.501 x 10
-1 Kr-85m 4.48 1.547 x 10 Activity Correction Formula E
No = Ne
)
where No = nuclide activity concentration at reactor shutdown N = measured nuclide activity concentration at time of sampling 1/2 = half life (hr)
A = In2/T
= decay constant (br-1) 1/2 l
t = time of sampling (in hours) after reactor shutdown l
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st1182-4113e165-12
STATION 9918.1F3 ENCLOSURE 5 ORIGINAL STATION 9918.1 ORIGINAL Page 1-CORE INVENTORY CORRECTION FACTORS Nuclide Effective Full Power Days of Coeration 4
30 60 90 120 150 180 210 280 310
-(1)
Kr-85m 1.0 Xe-133
.38
.98 1.0 Sr-91 1.0 Sr-92 1.0 I-131
.42
.94
.99 1.0 La-140
.41
.85
.97
.99 1.0 Cs-137 Use equation below (1),,,,, = 1.0 Cs-137 Correction Factor F =.0016 e +.50 where F = Cs-137 correction factor t = effective full power days Correction Factor EFPD
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Performed by:
s t1182-4113 f165-12 Date:
~
ENCLOSURE 6 STATION 9918.1 ~
CRIGINAL PAGE t FIGURE 1 Core Exit Fluid Temperature For Intdequate Core Cooling 1
2600 hW i
RC f
RC SUBCOOLED SUPERHEATED h3 2400
%D
=zsp 2200 b
T CLAD n===
2000
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REGION 3 T - c+F REGION 1 / REGION 2 3
1800 RR' m&
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CLAD >
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w v.
w CORE EKIT THERMOCOUPLE TEMPERATURE (F)
FIGURE 2 ENCLCSURE 6 STATICN 9918.1 CASE I - NORMAL OPERATION ORIGINAL PAGE 2 EXPECTED l-13'1 ACTIVITY CONCENTRATION in REACTOR COOLANT vs PERCENT FAILED FUEL 100
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ENCLOSURE 6 STABCN 9918.1 -
FIGURE 3 CRIGINAL PAGE 3 CASE 11 - OBSERVABLE MACROSCOPIC CLAD DAMAGE EXPECTED l-131 CONCENTRATION in REACTOR COOLANT vs PERCENT FAILED FUEL 10000 i
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ENCLOSURE 6 STATICN 9918.1 -
FIGURE 3 ORIGINAL PAGE 3 CASE !! - OBSERVABLE MACROSCOPIC CLAD DAMAGE EXPECTED l-131 CONCENTRATION in REACTOR COOLANT
~
vs PERCENT FAILED FUEL 10000 i,,,,
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FIGURE 4 ENCLOSUREC CASE 111 - SEVERE FUEL OVERHEATING STAECN 99 3.1 ORIGINAL
?^GE 4 EXPECTED NUCLIDE CONCENTRATION in REACTOR COOLANT vs PERCENT CORE OVERHEATED 4
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FIGURE 5 STATION 9918.1 ORIGINAL CASE Ill - SEVERE FUEL OVERHEATING PAGE 5 EXPECTED NUCLIDE CONCENTRATION in CONTAINMENT ATMOSPHERE vs PERCENT CORE OVERHEATED 0
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ENCLCSURE 6 STATION 9918.1 FIGURE 6 cpg 334g PAGE 6 CASE IV - FUEL MELTING EXPECTED NUCLIDE CONTAINMENT CONCENTRATION vs PERCENT CORE MELTED 0
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Fl.OWClh _ HETi10D H
- t l)ETERMINE DETERilINE CORE l>ETERMINE OPERA DETERMINE HOST STAR'l - 3PERAT10N REGION EXIT T/C AVERACE
_l)ETERNINE RECION USING CONDITION SEVERE OPERA l' ROM SPDS 8.2.la
'fEMPERATURE RCS PRESSURE ATOG 1)lACRAN 8.2.th B.2.fc B.2.1d lATTAINED - 8.2.ly I)ETERMINE CONT DETERMINE L)BTAIN PASS LNALYZE RCS SAH-y llYDROCEN
-CONTAINHENT
-SANPLES - RCS &
Pl.ES AND CORRECT p VOLDHE 't, RADIATION LEVEL SUNP LIQUID, r0 TlHE OF SUUT-4.2.If H.2. I g_
CONT CAS - 8.2.2a M)WN - 8.2.2h ANALYZE CONT ANAL.YZE CONT ATH SUNP SAtlPLE AND
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_.SAHPl.E Ahl) COR-p CORR TO TTHE OF RECT TO TlHE OF SHUTl)OWN - 8.2.26 311UTDOWN - 8.2.2d NORMAL. OPERATION KSTlHATE PERCENT_ ENI) 6.2.3a FAILEI) FUEL g
8.2.3a3 du flACROSCOPic KSTitlATE PERCENT
-CLAD DAMAGE FAILED FUEL ENI)
B.2.3b B.2.3h3 DETERMINE
'*%Y CORE
> CONDITION 8.2.3 55 VERE FUEL ESTitIATE PERCENT
_ l)VERilEATING
._. CORE OVERilEATED ENI 8.2.3c B.2.3c3 ESTlHATE PERCENT I
- FUEL tlELTING CORE HELTED END
(~D-B.2.3d 8.2.3d3 ENCI.0SURE 7 STATION 9918.1 ORIGINAL st1182-4113hl65-12 Page t
.