ML20004C516
| ML20004C516 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/20/1981 |
| From: | Geer T, Rosalyn Jones, Thornhill P BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20004C505 | List: |
| References | |
| RTR-NUREG-0565, RTR-NUREG-0737, RTR-NUREG-565, RTR-NUREG-737, TASK-1.C.1, TASK-2.K.3, TASK-2.K.3.30, TASK-TM TAC-44265, NUDOCS 8106040199 | |
| Download: ML20004C516 (20) | |
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p B&W'S POST TEST ANALYSIS FOR SEMISCALE TEST S-07-10D Si!a Doculaent No. 86-1125888-00 May 20, 1981
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F Principal Investigators T. E. Geer I
P. A. Thornhill R. C. Jones
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Prepared by i
Babcock & Wilcox Company
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for
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The Owners Group of Babcock & Wilco; 177 and 205 Fuel Assembly NSS Systems ll f3 i
8106040 M
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Introduction The United States Nuclear Regulatory Commission (NRC) has sponsored research and development programs related to a postulated loss-of-coolant W'
accident (LOCA) for light-water nuclear reactor systems. In order to j
evaluate the adequacy of the computer codes and models used in calculating transient behavior of the reactor coolant system during a small break LOCA, E
3; B&W was requested to provide a pretest prediction for the MOD-3 semiscale small break experiment (Test S-07-10B). The pretest prediction was com-pleted in October of 1979 and submitted to the NRC via Reference 1.
g Recently, the NRC requested that a post test evaluation of LOFT test q
L3-1 and semiscale test S-07-10D be performed. This report presents the post test analysis performed by B&W for the S-07-10D experiment. The L3-1 post test evaluation is presented in Reference 3.
The objectives of the EE post test analyses were outlined in Reference 2 to be:
1.
Evaluate the code predictive capability using initial and boundary conditions consistent with the actual test data.
2.
Identify code modifications and/or improvements necessary to q
predict the test data.
3.
Assess whether any improvements and/or modifications necessary for i
code predictions to agree with test data should be incorporated in-present ECCS small break evaluation models.
4.
Identify shortcomings in the test facility, instrumentation, etc.,
and their impact on code prediction capability, and recommend improvecents to the test facility, instrumentation, or test pro-gg cedures to improve the verification process.
A summary of this report is provided in Section 2.
A description of
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the semiscale system along with the relationship between the S-07-10B and S-07-10D tests is provided in Section 3.
Section 4 provides the analyses, results, and conclusions for the pretest and post test predictions.
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Summary & Conclusions _
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,j This report presents B&W's post test evaluation of Seniscale Test S07-10D.
In the pretest evaluation of S-07-10B, it was noted that the h[$
b; system pressure did not decrease to the ECCS actuation pressure. As part of that submittal, several potential causes for the overprediction were g:
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identified. Comparisons of the pretest prediction to the actual S-07-10D b!
test data generally confirmed that the causes identified in the pre test a
submittal were the sources for the discrepency between the prediction and u:*
the test data. To perform the post test analysis, input changes were made gg to eliminate identified sources of the discrepencies. As shown in Section 4 E
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substantial improvement was made in the prediction.
Relative to the specific concerns identified in Reference 2, the post test analysis confirmed that the CRAFT 2 code can predict the cmall break
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LOCA phenomenon observed in the test, provided that adequate test conditions are provided. No code modifications were necessary for the post test evalu-3 ation. However, core noding generally utilized for small break evaluations needed to be replaced by a more detailed and best estimate representa' tion.
This was necessary as the core noding used in the evaluation model results
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in conservative results when core uncovery occurs, as occurred in the S-07-10D test.
In order to improve the verification process, it is suggested that a
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similar a'pproach to that utilized for the recent L3-6 prediction be employed.
That approach consisted of setting up a " blind pretest" model, release of the test data, and then a post test evaluation wherein changes from the " blind pretest" model must be justified.
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S-07-10D Test l'
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Semiscale test S-07-10D represented a 10 percent cocmunicative cold leg 7:
' break at pump discharge, from a system initially at 2283 psi and 606 F (hot leg).
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$5 The experimental configuration is shown in Figure 1.
The primary coolant
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system has two loops; intact and broken. Each loop contains an operating 2
steam 8enerator and pump. A pressurizer is attached to the intact loop hot leg E3=
piping.
The reactor simulator consists of an electrically heated core, upper head, upper plenum and lower plenum. Total core power was 1.927 MW and core ff flow rate was 21 lbm/sec.
The broken loop steam generator steam valve was T@
open throughout the S-07-10D transient.
The B&W pretest prediction was based on test S-07-103; however, the 7;
initial conditions provided to B&W indicated that the broken loop steam generator secondary steam valve was open throughout the transient. Thus,
~2 the pretest model was set up with the valve open.
During the review of the S-07-10B test data, EG&G concluded that the steam valve did not remain open,
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but rather the valve actually closed at 17 seconds into the transient. To
- 3 compensate for this crror and provide appropriate test data for comparison, test S-07-10D was performed using initial system conditions sinilar to S-07-10B and with the broken loop steam generator steam valve left open.
Data for test
=
S-07-10D was submitted to B&W by References 4 and 5.
Since both B&W analyses, pretest and post test, were performed with the broken loop steam generator
,e steam valve assu 9d open, the applicable test data for cocparison purposes is that of the S-07-10D test.
%J, 4.
Analysis
!5
-4.1 Pretest Prediction m
R The B&W pretest analysis results and conclusions were submitted to the
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I NRC by Reference 1.
While the calculated results were decced to be reasonable
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's lE based upon the input assumptions utilized, it was recognized that the system h$
pressure was overpredicted as the actuation setpoint for the ECCS was not b"l reached. As part of the pretest submittal, potential causes for this over-NE
,7 prediction were identified. These were:
1.
Uncertainties in the blowdown of the broken loop generator.
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2.
Primary metal heat input appeared to be too large.
- 1 3.
The single node model of the core with all core heat being Es deposited into that node results in excess steam generation when the core starts uncovering.
(CORE 2AL option) f.0.~6 docs not 4.
The Bernoulli-Moody discharge model with Cd accurately predict leak flow when leak quality is high.
Experimental data for test S-07-10B was transmitted to B&W by Reference 6 on March 17, 1980.
Compsrison of the S-07-10B test data with the pretest pre-diction generally confirmed that the causes identified in the original pretest submittal were the causes for discrepancy between the analysis results and the a=
test data. As a result of this review, it was decided to make the following major model changes for the post test evaluation:
1.
Change the core model from a single node representation to several nodes to eliminate the conservatisms inherently associated with the CORE 2AL option.
2.
Force the broken loop steam generator secondary pressure to conform to the actual test pressure, since inadequate information is availa-ble to predict the secondary side response.
3.
Utilize the REM discharge model with Cd = 1.0 for two phase break flow.
"5 The post test analysis, described in detail in the following section,
. 2.;
demonstrates that these model changes substantially improve the predicted E.
results.
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4.2 Post Test Prediction f"
4.2.1 Post Test Analysis Model i
h.M The calculations performed and model developed for the pretest prediction F
p were used extensively in the post test analysis. Modifications made to the
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'S pretest model to obtain the post test model, along with a description of the
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-approach taksn, are presented in the following paragraphs.
'l As shown in Reference 5, the initial conditions for tests S-07-10B and S-07-10D vere essentially the same; thus, the initial conditions used for the
- .;r Pratest prediction were deemed to be valid for the post test prediction.
13 Table 1 contains both sets of initial conditions. The ECC parameters, however, 5
had significant variation between the two tests and the correct values were
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input for the post test analysis. Table 1 also provides the ECC parameters used in the S-07-10D analysis.
The post test prediction was performed using version 17.0 of the CRAFI2 computer code (ref. 7), rather than version 8.4.
Version 17.0 of CRAFT 2 makes available the HEM model for break discharge.
The computer deck was 5
f derived from the pretest prediction deck with changes made as required by format changes between version 8.4 and 17.0.
As noted in Section 4.1, there was uncertainty in the steam generator s
blowdown calculated in the pretest analysis; There were two factors which preclude an accurate prediction of steam generator secondary blowdown through 5
the open steam valve:
1.
Information on the secondary system piping geometry was not available.
2.
Information on fluid quality leaving the secondary system was not available, due to uncertainty on the efficiency of the steam g
separator used in the test.
Lacking this information, it is very difficult to accurately predict steam i
flow characteristics.
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- To compensate for these unknowns, the pressure from the broken 1 cop t
i si steam generator secondary was input as a boundary condition in the post test model. This was accomplished by utilizing the CRAFT relief valve actuation
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6 h,
E pressure versus time table.
Since beth -he intact loop and broken loop j
steam generators would be forced to follow the same pressure versus time table, heat transfer in the intact loop steam generator was stopped at ff 100 seconds by setting the heat transfer coefficient equal to zero. This action was judged to be acceptable because it had been determined (Ref-
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erence 5 page 11) that the primary and secondary systees were somewhat decoupled during the majority of the transient due to high void fractions
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in the steam generator tubes and resultant poor heat transfer.
a As was orginally noted in the pretest submitt.al, primry metal heat input appeared to be too large in the pretest prediction. Since no pri-
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E mary metal tesperature histories are available, a detailed evaluation of the primary metal heating concern could not be performed. However, at
- j
this time it is believed that the primary metal heats were reasonable.
In the pretest prediction for test S-07-10B, the CORE 2AL option of the CRAFT computer code was utilized. With the option all core heat, regardless of flow direction, is deposited in a specified node. In the
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q pretest prediction all core heat was deposited in the single core node.
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When the code predicts substantial core uncovery, as occurred in the S-07-10B rs analysis, use of the CORE 2AL option produces extremely conservative results relative to core uncovery.
This occurs because the core heat in the uncovered E
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portion of the rods, is placed into the remaining liquid in the core. This leads to excessive boiling in the remaining liquid, and thus excessive pres-N surization. This assumption is conservative for.small break LOCA predictions.
f In the post test analysis, the CORE 2AL option was eliminated.
In order to allow for the heater rod axial power distribution, the core was modeled as several nodes.
By doing this, only a small portion of the energy was added s
to the flow paths at the top and bottom of the core and most of the energy was added to the central portion of the core.
By using a cultinode representation,
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the CRAFI calculated core heat transfer coef ficients would now be based upon.
W the inlet fluid condition to the core flow path. Thus, during a core uncovery h
situation, a low surfar,e heat transfer coefficient would be chosen for.the If core paths which have steam inlet conditions. In this manner, the actual heat transfer conditions in the core, i.e. good heat transfer in the covered 3,
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M portion of the rod vis-a-vis poor heat transfer in the uncovered portion of gi the rod, would be more closely approximated than by the CORE 2AL, single node q,
core representation.
In the pretest prediction the orifice equation and the Moody model, respectively, were utilized for subcooled and two-phase discharge models.
Both models utilized a discharge coefficient of 0.6, which resulted in an
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underprediction of break flow for two-phase and steam discharge. For the post test analysis, the orifice equation with discharge coefficient equal to 0.6 was utilized for the subcooled condition, and the 11EM model with discharge coef ficient equal to 1.0 was used for two-phase fluid and steam U
conditions.
The resultant post test evaluation noding diagram is illustrated in
- n Figure 2.
In summary, the differences between the post test and pretest models are:
1.
The ECC parameters for the post test evaluation were changed to represent the actual S-07-10D test condition.
2.
The steam generator secondary pressure for the broken loop was used as a boundary condition for the post test evaluation. Due p
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to code limitations, the intact loop steam generator secondary pressure had to be equivalent to that of the broken loop steam generator. To minimize the feedback of the intact loop steam
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generator pressure, the heat transfer coefficient was set to zero f~
after 100 seconds, t
3.
A multinode core representation (6 nodes /7 paths) was chosen for jd the post test evaluation.
N 4.
The saturated fluid discharge model was changed from the Moody
. correlation with a CD = 0.6 to the llEM correlation with a CD = 1.0.
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4.2.2 Post Test Results r:n
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A sequence of events comparison between the experimental data and the post test analysis is provided on Table 2.,
For reference purposes, the il In general, a substantial improve-I
.9) pretest evaluation values are also shown.
ment in resul'ts were obtained with the post test analysis, especia1.y as
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related to core level response.
Figures 3 through 7 present results of the
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evaluation. These are discussed more fully below.
l,j,3 Figure 3 presents a comparison of system pressure for both the pretest le E5 and post test results with the S-07-10D test pressure. The post test result
"!r is very close to the test data. Especially of interest is that the post f~
test curve did not flatten out after 250 seconds like the pretest did. Tha HPIS came on at 467 seconds in the post test prediction as compared to 460
- F s
seconds in.the test. The HPIS did not actuate in the pretest per. diction.
- er.
g Examination of Figure 4 shows the reason for the differences.
rigure 4 presents a comparison of the integrated net energy removed from the primary system for the pretest and post test predictions. Energy
- f into the primary systa.m was from the core and primary metal, and er.ergy out of the primary system was due to break flow and heat transfer to the steam generator secondaries.
As is seen from the figure, after approximately 200 t
l-seconds there was little net energy removed fr.om the system for the pretest
' Hi prediction. However, the revised core and break flow models result in a
- w
- p continuous net energy removal and depressurization.
Figure 5 presents pressure versus time for the steam generator secondaries.
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- As was previously noted, the secondary side pressure for both the intact and f
broken loop was forced to follow actual broken loop test pressure, during j.
the post test -prediction. Thus, the broken loop steam generator blowdown 5
has been removed as a potential cause of errors in the post test analysis.
=
5) 9 b5 Figure 6 presents break mass flow rate versus time for the test, I
pretest prediction and post test prediction. It can be seen that the
- 5 HEM discharge model with discharge coefficient of 1.0 (used in the
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post test prediction) provided results in closer agreement than the pretest prediction. However, the predicted mass flow rate was still g
5 somewhat b, low the actual test data. This is generally a result of g
the lower system pressure which was predicted as compared to the actual data (Figure 3).
r Figure 7 is a presentation of core collapsed liquid level for the 3;;
test, the post test prediction and the pretest prediction. As can be seen, the post test result compares favorably with the test data and is considerably different than the pretest prediction. This large change is attributed to a co=bination of the correction in the steam generator blowdown and the cultinode core model.
It is important to note that vessel refill, that occurred at =100 seconda due to the loop seal clear-ing, was calculated to occur in the post test prediction. Previously, none of the semiscale parcicipants had predicted the vessel refill as
- was noted in Reference 5.
/
4.3 Conclusions In general the post test prediction, through the time of HPIS actua-tion, was quite satisfactory and in good agreement with S-07-10D test data.
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Relative to the questions of Reference 2, the following specific conclusions can be m.3de:
79 1.
The post test evaluation, using initial and boundary conditions y
li consistent with the actual S-07-10D test data, shows good agree-
'G ment with test results.
Jg 2.
No computer code codifications or improvements were found
$1' necessary to preoict the test data. However, the CORE 2AL l
option normally used in the small break LOCA evaluation model 1
I 10
,i l
1 - ;
needed to be replaced by a more detailed core representation to obtain best estimate calculations.
3.
The CORE 2AL option is conservative for the evaluation model; thus, the detailed changes made to obtain best estimate results need not be incorporated into the evaluation model.
4.
Better specification of test parameters is necessary before calculations are performed.
It is suggested that any future p
analysis be perfgrmed in the post test mode, similar to the p,
L3-6 prediction 5.
References Letter from J. H. Taylor (B&W) to Richard P. Denise (NRC),
1.
" Analysis Prediction for Test S-07-10B", dated October 9, 1979.
2.
Letter to al?. B&W Licensees from Robert W. Reid (NRC), dated February 24, 1981.
3.
N. K. Savani, R. C. Jones, " B&W's Post Test Evaluation of LOFT Test
- 3-1," Doc. No. 51-1125988-00, May 1981.
4.
D. J. Shimeck, " Analysis of Semiscale MOD-3 Small Break Tests S-07-10 and S-07-10D", (EG&G Report E3G-SEMI-5201), July 1980.
5.
Letter to J. H. Taylor (B&W Licensing) from J. A. Dearier (Code Assesscent and Applications Branch) " Transmittal of Small Break Experiment Preliminary Comparison Report to Participants - JAD-4-81" dated January 8, 1981.
6.
Letter from L. E. Phillips (Division of Systems Safety) to r:.
J.11. Taylor (B&W Licensing), " Experiment Data Delease -
[
Semiscale Test S-07-10B", dated March 17, 1980.
7.
R. A. Hedrick, J. J. Cudlin and R. C. Foltz, CRAFI2 - Fortran y
Program for Digital Simulation of a Multinode Reactor Plant F
During Loss of Coolant, BAW-10092P, Revision 2, Pabcock & Wilcox, April 1975.
bp 8.
J. R. Paljug, M. D. Charakhani, R. C. Jones, B&W's Best Estimate Prediction of the LOFT L3-6 Nuclear Small Break Test Using the p
CR'Fr2 Computer Code, Babcock & Wilcox, March 1981. Transmitted
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vl.t letter from J. H. Taylor (B&W) to P. Check (NRC), March 20, 1981.
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- 5 Table 1.
Initial Conditions and ECC Requirements for Semiscale Test S-07-10D
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(E Value Used Actual Data 2'
Initial Cond.4.tions from S-07-10B for S-07-10D Nominal System Pressure, psia 2250 2277 Hot Leg Fluid Temperature, F
604.4 605 15 Cold Leg Fluid Temperature, F
535.8 541 Core AT, 'F 68.6 63 Core Inlet Flow, Lbc/sec.
21.4 21 Total Core Power, g; 1.927 1.94 ECC Paraceters Intact Loop Accumulator (Flood Tank)
System Pressure at Actuation 232 psia Tank Pressure at Actuation 450 psia Liquid Volume 1.6 fc.3 0.g8ft.3 Cas Volume Temperature 80 F E
Intact Loop HPIS Actuation Pressure 232 psia Injection Rate (Average) 0.17 Lbm/sec.
?Ei Temperature 80 F t
Intact Loop LPIS Actuation Pressure 305 psia Injection Rate (Average) 0.24 Lbm/sec.
Temperature 80 F 5u
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Table 2.
Sequence of Events Test S-07-10D Pretest Prediction Pcat Test Prediction Event Time (sec.)
Time (sec.)
Time (sec.)
Blowdown initiated 0
0 Pressurizer pressure = 1800 psia 6.9 5.65 5.65 Begin core power decay 7.7 5-10 Pump coastdown initiated Upper plen 2m fluid saturates 8.0 6-10
.5-10 Pressurizer empties 20 20
=15 Entire system saturated 27
, 60 35-40 Upper plenum liquid level reaches hot leg 42 25 Pressure suppression system pressure reduction begins 52 60 Liquid from cold legs drains to vessel and pump suctions resulting in two-phase mixture at break 65-90
=60 40-55 Power to pump terminated 69 69.6 69.6 Pumps stop 79
=80 Top of support tubes uncovered in upper head 80 80 Pressure suppression system tank pressure reduction finished 160 250 Start dryout of core 268 260 Core completely voided 434 330 Lowest point in post test (93%)
467 Accumulator injection begins 460 470 HPIS injection begins 460 467 9
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MOD-3 SEMISCALE
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COLD LEG BREAK ASSEMBLY
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4 t(.l Figuro 2 N00AL DIAGRAM FOR SEMISCALE S0710D POSTEST PREDICTIONS BROKEN LOOP INTACT LOOP 85 20 ii 84 40 A
28 41 19 PZR 69 70 67 68 71 72 73 74 25 9
31 20 18 G1 43 142 39l 13 22 l l 2126l 125 44 34 6
21 17 16 7
8 12 45 37 35 II 19 16 8
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Figure 7 CORE COLLAPSED LIQUID LEVEL 13.5 1
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S-07-100 Ig POST TEST h
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1 ESTIMATED UNCERTAINTY IN CORE LIQUID LEVEL 10.0 1
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