ML20010G862
| ML20010G862 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/09/1981 |
| From: | Geer T, Rosalyn Jones, Thornhill P BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20010G847 | List: |
| References | |
| TASK-2.K.3.30, TASK-TM 86-1125888-01, 86-1125888-01-0, 86-1125888-1, TAC-45817, NUDOCS 8109220600 | |
| Download: ML20010G862 (24) | |
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B&W'S POST TEST ANALYSIS FOR SEMISCALE TEST S-07-10D Doctanent No. 86-1125888-01 June 9, 1981 lI I
Principal Investigators
>E T. E. Geer W
P. A. Thornhill R. C. Jones Prepared by Babcock & Wilcox Company for The Owners Group of Bebcock & Wilcox 177 and 205 Fuel Assembly NSS Systems
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I I!!TRODUCTION 1
The United States NJclear Regulatory Commission (NRC) has sponsored research and development programs related to a postulated 70:s-of-coolant accident (LOCA) for light-water nuclear reactor systems.
In order to evaluate the adequacy of the computer codes and models used in calculating transient behavior of the reactor coolant system during a small break LOCA, B&W was retiuested to provide a pretest prediction for the M00-3 semiscale small break experiment (Test S-07-108). The pretest prediction was completed in October of 1979 and submitted to the NRC via Reference 1.
I Recently, the NRC requested that a post test evaluation of LOFT Test L 3-1 and Semiscale Test S-07-10D be performed.
This report presents the post test anclysie performed by B&W for the S-07-100 experiment. The L3-1 post test evaluation is presented in Refere :e 3.
The objectives of the I
post test analyses were outlined in Refer?nca 2:
I Evaluate the code predictive capability using initial and boundary 1.
conditions consistent with the actual test data, 2.
Identify code modifications and/or improvements necessary to predict the test data, i
3.
Assess whether any improvements and/or modifications necessary for code a
predictions to agree with test data should be incorporated in present ECCS small break evaluatior, models, 4.
Identify shortcomings in the test f acility, instrumentation, i:tc.
and their impact on code prediction capability, and recommend improvements to the test facility, instrumentation, or test procedures to improve the verification process.
A summary of this report is provided in Section 2.
A 1escription of the cemiscale system along with the relationship between the 5-07-108 and S-07-100 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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2.
SUMMARY
& CONCLUSIO.,5 This report presents B&W's post test evaluation of Semiscale Tes't S-07-100.
In the pretest evaluation of S-07-108, it was noted that the system pressure did not decrease M the ECCS actuation pressure.
As part of that submittal, several potential causes for the overprediction were identified.
Comparisons nf the pretest prediction to the actuai S-07-10D test data generally confirmed that the causes identified in the pretest submittal were i
the sources for the discrepancy between the prediction and the test data.
To perform the post test analysis, input changes were made to eliminate identified source of the discreps.;cies. As shown in Section 4, substantial improvement was made in the prediction.
I Relative to the specific concerns identified 5 Reference 2, the post te:.t I
analysis confirced that the CRAFT 2 code car. predict the small break LOCA g
phenomenon observed in the test, provided ti.ac adequate test conditions are provided.
No code modifications were necessary for the post test i
i evaluation.
Ilowever, core noding generally utilized for small break I
evaluations needed to be replaced by a more detailed and best estimate representation.
This was necessary as the core noding used in the eva'uation model results in conservative results when core uncovery occurs, as occurred in the S-07-100 test.
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3.
S-07-10D TEST I
5emiscale Test S-07-10D represented a 10 percent communicative cold leg break at the pump discharge, from a system initially at 2263 psig and 606*F l
(hot leg). The experime< ul configuration is shown in rigure 1.
The primary coolant system ha two loops, intact and broken.
Each loop contains an operating steam generator and pump. A pressurizer is attached to the l
intact loop hot leg piping. The reactor simulator consists of an elcetrically heated core, upper head, upper plenum and lower plenum.
Total core power was 1.927 W and cora flow rate was 21 lbm/sec.
The broken loop steam generator steam valve was open throughout the S-07-100 transient.
The B&W pretest prediction was based on test S-07-108: however, the init h'i conditions provided to B&W indicated that the c.'oken loop steam generator secondary steam valve was open throughout the transient.
Thus, 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, but rather the valve actually closed at 17 seconds into the transient.
To l
compensate for this error and provide appropriate test data for comparison, l
Test S-07-10D was performed using initial system conditions similar to S-07-10B and with the broken loop steam generator steam valve left open.
Data for t.2 Test S-07-100 was submitted to B&W by Reference 4 and 5.
Since both B&W analyses, pretest and post test, were performed with the broken
- I loop steam generator steam valve assumed open, the applicable test data for comparison purposes is that of the S-07-10D test.
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ANALYSIS 4.1 Pretest Prediction The B&W pretest analy!i', results and conclusions were submitted to the NRC by Reference 1.
While the calculated results were deemed to be reasonable based upon the input assumptions utilized, it was recognized that the system pressure was overpredicted as the actuation setpoint for the ECCS was not reached.
As part of the pretest submittal, potential causes for this overprediction were identified.
These were:
I 1.
Uncertainties in the blowdown of the broken loop generator, 2.
Primrey me.al heat li put appeared to be too large.
3.
The Wgle node model of the core with all core heat being E
deposited into that node results in excess steam generation when
'E the core starts uncovering, (CORE 2AL option) 4.
The Bernouli?-Moody discharge model with C4 of 0,6 does not I
accurately predict leak flow when leak quesity !s high.
Experimental data for Test S-07-10B was trasmitted to B&W by Reference 6 on March 17, 1980.
Comparison of the S-07-10B test data with the pretest prediction generally confirmed that the causes identified in the original pretest submittal were the causes for l
discrepancy between the analysis results and the test data.
As a result of this review, it was decided to make the following major model changes for the post test evaluation:
1.
Changc the core model frw a single node representati in to several I
nodes to eliminate the conservatisms inherently associated with the CORE 2AL option.
2.
Force the broken loop steam generator secondary pressure to conform I
to the actual test pressure, since inadequate information is available to predict t.1e secondary side response.
3.
Utilize the Homogeneous Equilibrium Model (HEM) discharge nodel j
with Cd = 1.0 for two phase break flow.
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The post test analysis, described in detail in the following section, demonstrates that these model changes substantially improve the predicted results.,
4.2 Post Tes:. Prediction 4.2.1 Post Test Analysis Model The calculations performed and rt: del developed for the pretett prediction were used extensively in the post test analysis.
4 Modifications made to the pretest model to obtain the post test l
l model, along with a description of the approach taken, are presented in the icllowing paragraphs.
1 As shown in Refere*e 5, the initial conditions for Tests l
S-07-10B and S-Di-10D were essentially the same; thus, the initial conJitions used for the pretest prediction were deemed to be valid for the post test prediction. Table 1 contains both I
sets of initial conditions.
The ECC parameters, hcwever, had significant variation between the two tests and the correct values were input for the post test analysis.
Table 1 also provides the ECC parameters used in the S-07-10D analysis.
I The post test prediction was performed using version 17.0 of the CRAFT 2 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 derived from the pretest prediction deck with chenges made as required by format changes bd ween version 8.4 and 17.0.
I As noted in Section 4.1, there was uncertainty in the steam generator blowdown alculated in the pretest analysis.
There were two factcr s which preclude an accurate prediction of steam generator secondary blowdown through the open steam valve:
1.
Information on the secondary system piping geometry was nGt, available.
2.
Information on fluid qJality leaving the secondary system was not available, due to uncertainty on the efficiency of the steam separator used in the test. i i
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I Lacking this biet mation. R is very difficult to accurately predict steam flow characteristics.
To compensate f these unknowns, the pressure from the breken loop steam generator secondary was input as a boundary condition in the post test model. This was accomplished by utilizing the CRAFT relief valve actuation pressure versus time table.
dnce both the intact loop and broken loop steam generators would be forced to follow the sama pressure versus tir.r table, heat transfer in tie. intact loop steam generator was stopped et 100 seconds by setting the heat transfer coefficient equal i ero.
This action was judged to be acceotable because it had been i
determined that the primary and secondary systems were somewhat decoupled during the majority of the transient due to high void fractions in the steam generator tubes and resultant poor heat transfer (Refererre 5, Page 11).
It was originally noted in the pretest submittal that primary metal heat input appeared to be too large.
Since nc primary metal temperature histories are available, a detailed evaluation of the primary metal :. eating concern could not be performed.
However, at thin time it is believed that the primary metal heats were reasonable.
8 In the pretest predictiun for Test S-07-1Cd, the CORE 2AL option of the CRAFT computer code was utilized.
In the aretest prediction all core heat regardless of flow direction was deposited in the u.acle core node. When the code predicts substantial core uncovery, as occurr;u in the S-07-10B analysis, use of the CORE 2f'. option produces extremely conservative results relative to core uncovery because the core heat in the uncovered portion of the rods is olaced into the remai..ing _
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r liquid in the core. This leads to ucessive boiling in the remaining liquid, and thus excessive pressurization.
This assumption is conservative for small break LOCA predictions.
I In the post test analysis, the CORE 2AL option was eliminated.
In ceder to allow for the heater rod axial power distribution, the core was modeled as several nodes.
By doing this. only a i
small portion of the energy was added 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. Py using a multinode representation, the CRAFT calculated core heat transfer coefficients would now be bast.J mon the inlet fluid condition to the core flow path.
Thus, during a core uncovery situation, g
a low surface heat transfer coefficient would be chosen for the 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 portion of the rod vis-a-vis poor heat transfer in the uncovered portion of the rod) would be more closely approximated than by the CORE 2AL, single node core representation.
In the pretest prediction the orific.e equation and the Moody mcdel, respectively, were utilized for subcooled and two-phase discharge models. Both models utilized a discharge coefficient of 0.6, which resulted in an underprediction of break flow for two-phase and steam discharge.
For the post test analysis, the i
orifice equation with discharge coefficient equal to 0.6 was utilized for the subcooled condition, and the HEti model with dircharge coefficient equal to 1.0 was used for two-phase fluid and steam conditions.
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The resultant post test evaluation noding diagram is illustrated in figure 2.
In sumnary, the differences between the post test and pretest models are:
I 1.
The ECC parameters for the post test evaluation were changed to represent the actual S-07-100 test condition.
2.
The steam generator secondary pressure for the broken loop I
was used as a boundary condition for the post test evaluation.
Due to code limitations, the intact loop steam generator secondary pressure had to be equivalent to that I
of the broken loop steam generator. To minimize the feedback of the intact loop steam generator pressure, the heat transfer coefficient was set to zero after 100 seconds.
3.
A multinode core representation (6 nodes /7 reaths) was chosen for the post test evaluation.
I 4.
The saturated fluid discharge model was changed from the Moody correlation with a Cq = 0.6 to the HEM correlation with a Cd = 1.0.
4.2.2 Post Test Results A sequence of events comparison between the experimental data and the post test analysis is provided on Table 2.
For reference purposes, the pretest evaluation values are also shown.
In general, a substantial improvement in results were obtained with the pod test analysis, especially as related to i
core level response.
Figures 3 through 7 present results of the evaluation.
These are discussed more fully below.
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Figure 3 presents a comparison of system pressure for both the pretest and post test results with the S-07-10D test pressure.
The post test result is very close to the test data.
Especially of interest is that the post test curve did not flatten out lI after 250 seconds like the pretest did.
The HPIS came on at 467 l
seconds in the post test prediction as compared to 460 seconds j
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5 in the test. The bgh pressure injectian system did not actuate 1.1 the pretest prediction.
Examination of Figure 4 shows the reason for the differences.
I Figure 4 presents a comparison of the integrated net energy removed from the primary system for the pretest and post test predictions.
Energy into the primary system was from the core and primary metal, and energy out of the primary system was due to break flow and heat transfer to the steam generator.
Figure 4 illustrates that after approximately 200 seconds there was little net energy removed from the system for the pretest prediction.
However, the revised core and break flow models result in a continuous net energy removal and depressurization.
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.tigure 5 presents pressure versus time for the secondary side of the steam generater. As was previously noted, the secondary side pressure for both the isrcact and broken loop was forced to i
follow actual broken loop test pressure, during the post test prediction. Thus, the broken loop steam generator blowdown has been removed as a potential cause of errors in the post test analysis.
I Figure 6 presents break mass flow rate versus time for the test, pretest prediction, and pos+ test prediction.
It can be seen that the HEM discharge model with discharge coefficient of 1.0 (used in the post test prediction) provided results in closer I
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the pretast prediction. However the predicted mass flow rate was still somewhat below the actual test data. This is geners11y a result of the lower system pressure which was predicted as compared to the actual data (Finure 3).
Figure 7 is a presentation of core collapsed liquid level for the 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 combination of the correction in the steam generator blowdown and the multinode core model.
It is important t.
note that vessel refill, that occurred at *100 seconds due to the loop seal clearing, was calculated to occur in the post test prediction.
Previously, none of the semiscale participants had predicted the vessel i
refill as was noted in Reference 5.
4.3 Conclusions In general the post test prediction, through the time of HPIS actuation, was quite satisfactory and in good agreement with S-07-10D test data. Relative to the questions of Reference 2, the following I
specific cor:lusions can be made:
1.
The post test evaluation, using initial and boundary conditions consistent with the actual S-07-100 test data, shows good agree-ment with test results.
2.
No computer code modifications or improvements were found necessary to predict the test data.
However, the CORE 2AL optiun normally used in the small break LCCA evaluatien model needed to I
be replaced by a more detailed core representation to obtain best estimate calculations.
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The CORE 2AL option is conservative for the evaluation model; thus, the detailed changes made to obtain best estimate results I
need not be incorporated into the evaluation.nadel.
1 4.
Better specification of test parameters is necessary before calculations are performed.
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REFERENCES 1.
Letter from J. H. Taylor (B&W) to Mchard P. Denise (NRC), " Analysis Prediction for Test S-07-10B", datd October 9, 1979.
2.
Letter to all B&W Licensees from Robert W. Reid (NRC), oated February 24, 1981.
3.
N. K. Savani, R. C. Jones, "B&W's Post Test Evaluation of LOFT Test L3-1" Doc. No. 51-1125988-00, May 'E1.
4.
D. J. Se
'ck, " Analysis of Semiscale MOD-3 Small Break Tests S-07-10 and S-07-10D", (EG&G Report EGG-SEMI-5201), July 1980.
5.
Letter to J. H. Taylor (B&W Licensing) from J. A. Dearier (Code Assessment and Applications L*.nch NRC) " Transmittal of Small Break Experiment Preliminary Comparison Report to Participants - JAD-4-81",
I dated January 8, 199.
6.
Letter from L. E. Phillips (Divisica of Systems Safety) to J. H. Taylor I
(B&W Licensina), " Experiment Data Release - Semiscale Test S-07-10B",
dated March 1,, 1980.
7.
R. A Hedrick, J. J. Cudlin and R. C. Foltz. CRAFT 2 - Fortran Program
,I for Digital Simulation of a Multinode Reactor Plant During tess of Coolant, BAW '0092P, Revision 2, Babcock & Wilcox, April 1973.
I 8.
J. R. Paljug, M. D. Gharak'.ri, R. C. Jones, B&W's Best Estimate Prediction of the LOFT L3 6 ft lear Small Break Test using the CRAFT 2 Computer Code, Babcock & Wilcox, March 1981.
Transmitted via letter from J. H. Taylor (B&W) to P. Check (NRC), March 20, 1981.
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I Table 1.
Initial Conditions and i'cC Requirements for Semiscale Test S-07-10D I
Value Used Actual Data Initial Conditions from S-07-105 for S-07-10D I
fiominal System Pressure, psia 2250 2277 Ilot Leg Fluid Temperature, F
604.4 605 Cold Leg Fluid Temperature, F
535.8 541 Core AT, F
68.6 63 l
Core Inlet Flow, Lbm/ soc.
21.4 21 Total Core Power, FM 1.927 1.94 I
ECC Parameters Intact Loop Accunulator (Flood Tank)
I System Pressure at Actuation 232 psia Tank Pressure at Actuation 450 psia Liquid Volume 1.6 ft.3 0.g8ft.3 I
Cas Volume Temperature 80 F Intact Loop IIPIS Actuation Pressure 232 psia I
Injection Rate (Average) 0.17 Lbm/sec.
Temperature 80 F I
Intact Rop LPIS Actuation Pressure 305 psia
'I Injection Rate (Average) 0.24 Lbm/sec.
Temperature 80 F
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Sequence of Events Test S-07-10D Pretest Prediction Post Test Prediction Event Time (sec.)
Tity (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 plenum fluid saturates 8.0 6-10 5-10 Pressurizer emptics 20 20
=15 Entire system saturated 27 60 35-40 Upper plenum liquid icvel 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 Pcwer to pump teruinated 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 (935) 467 Accumulator injection begins 460 470 IIPIS injection begins 460 467
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