ML20010B130
| ML20010B130 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/10/1981 |
| From: | Hukill H METROPOLITAN EDISON CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| L1L-221, NUDOCS 8108140061 | |
| Download: ML20010B130 (2) | |
Text
.s Metropolitan Edison Company Post Office Box 480 kI
-L Middletown, Pennsylvania 17057 Writer's Direct Dial Number August 10, 1981 LlL 221 G)
$1l(
$\\s Office of Nuclear Reactor Regulation g
hh Attn: John F. Stolz, Chief r
AUGy "o O,OlA N2 Operating Reactors Branch No. 4
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U. S. Nuclear Regulatory Commission y,,,
.coMf' gross Washington, D.C.
20555
Dear Sir:
^
N to Three Mile Island Nuclear Station, Unit ik
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Operating License No. DPR-50 Docket No. 50-289 Small Break LOCA Model Enclosed please find a copy of "B&W's Post Test Analysis for Semi-Scale Test S-07-10D" dated June 9, 1981. This document together with our response of July 2, 1979 (TLL 197) (Loft Test L3-1) provides response to your letter of February 24, 1981. The results are summarized below:
1.
The evaluations provided demonstrate that the present small
~
break analysis techniques result in good agreement with the test data when actual test conditfans are consideced.
2.
No code modification and/or improvements were necessary to predict the experiments. However, as noted in the S-07-10D post test analysis, a more detailed core representation was necessary to provide a best estimate simulation of the ex-periment due to the extensive core uncovery whica occurred in the test.
3.
The core representation in the Evaluation Model gives con-servative results when core uncovery occurs. The more detailed representation of the core need not be included, therefore, in the Evaluation Model.
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8108140061 910810 PDR ADOCK 05000289 P
PDR Metroookton Edison Company is a Member of the Genera l Pubhc Utihties System
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, 4.
In order to improve the verification process, it is suggested that an approach similar to that utilized for.the recent L3 prediction be employed.. That approach consists of setting up and submitting a " blind pre-test" model. Then following re-ceipt of the test data, perform a post test evaluation and resubmit the results together with justifications for any model changes from the " blind pre-test" model.
Sincerely, b
i N
H. D. Hukill Director, TMI-l cc:
B. H. Grier (w/o enclosure)
L. Barrett (w/o enclosure)
R. Jacobs (w/o enclosure)
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B&W'S POST TEST ANALYSIS FOR SEMISCALE TEST S-07-10D Document No. 86-1125888-01 June 9, 1981 5
5 Principal Investigato'rs T. E. Geer I
P. A. Thornhill R. C. Jones I
Prepared by Babcock & Wilcox Company for The Owners Group of Babcock & Wilcox 177 and 205 Fuel Assembly NSS Systems I
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!'I 1.
INTRODUCTION The United States Nuclear Regulatory Commission (NRC) has <.ponsored research and development programs related to a postulated loss-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 requested to provide a pre'sst prediction for the MOD-3 semiscale small break expariment (Test S-07-10B). The pretest prediction was completed in October of 1079 and submitted to the NRC via Reference 1.
Recently, the NRC requested that a post test evaluation of LOFT Test L3-1 and Semiscale Test S-07-100 be performed.
This report presents the post i
test analysis performed by B&W for the S-07-103 experiment.
The L3-1 post test evaluation is presented in Reference 3.
The objectives of the I
I post test analyses were outlined in Reference 2:
1.
Evaluate the code predictive capability using initial rad boundary conditions consistent with the actual test data, 2.
Identify code modifications and/or improvements necessary to predict the I
test data, 3.
Assess whether any improvements and/or modifications necessary for code 8
predictions to agree with tast data should be incorporated in present ECCS small break evaluation models, 8
4.
Identify shortcomings in the test facility instrumentation, etc., 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 description of the semiscale system along with the relationship between the S-07-108 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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2.
SUMMARY
& CONCLUSIONS This report presents B&W's post test evaluation of Semiscale Test S-07-100.
In the prei.est evaluation of S-07-10B, it was no+.ed that the system pressure did not decrease to the ECCS actuation pressure.
As part of that submittal, several potential causes for the overprediction were identified.
Comparisons of the pretest prediction to the actual S-07-100 test data generally confirmed that the causes identified in the pretest submittal were the sources for the discrepancy between th3 prediction and the test data.
To perform '.he post test analysis, input changes were made to eliminate identified sources of the discrepancies. As shown in Section 4, substantial improvement was made in the prediction.
l Relative to the specific concerns identified in Reference 2, the past test 8
analysis confirmed that the CRAFT 2 code can predict the small break LOCA phenomenon observed in the test, provided that adequate test conditions are I
provided. No code modifications were necessary for the post test evaluation. However, 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 evaluation model results in co:servative results when core uncovery occurs, as occurred in the S-07-100 test.
In order to improve the verification process, it is suggested that a similar approach to that utilized for the recent L3-6 prediction be employed. That approach consisted of setting up a "blf nd pretest" model, release of the test data, and then a post test evaluation wherein chang 2s from the " blind pretest" model must be justified.
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7 S-07-100 TES1 3
Semiscale Test S-07-10D represented a 10 percent communicative cold leg i
break at the pump discharge, from a system initially at 2283 psig and 606*F l
l (hot leg). The experimental configuration is shown in Figure 1.
The primary coolant system has two loops, intact and broken.
Each loop contains j
an operating steam generator and pump. A pressurizer is attached to the 1
intact loop hot leg piping.
The reactor simulator consists of an electrically heated core, upper head, upper plenum and lower plenun. Total core power was 1.927 MW and ccre flow rate was 21 lbm/sec.
The broken loop steam generator steam valve was open throughout the S-07-10D transient.
I The B&W mitest prediction was based on test S-07-10B: hor',er, the initial conditions provided to d&W indicated that the broken loop steam generator secondary stean valve was open throughout the transient. Thus, the pretest 4
nodel 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 compensate for this error and provide appropriate test data for comparison, Test S-07-10D was performed using initial system conditions similar to S-07-108 and with the broken loop steam generator steam valva left open.
Data for the Test S-07-100 was submitted to C&W by Reference 4 and 5.
Since both B&W analyses, pretest and post test, were performed with the broken 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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I 4.
ANALYSIS 4.1 Pretest Prediction The B&W pretest analysis 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 satpoint for the ECCS was not reached. As part of the pretest submittal, potential causes for this overprediction were identified.
These were:
1.
Uncertainties in the blowdown of.he broken loop generator, 2.
Primary nietal heat input appeared to be too large.
3.
The single node model of the core with all core heat being deposited into that node results in excess steam generation when i
I the core starts uncovering, (CORE 2AL option) 4.
The Bernoulli-Moody discharge model with C of 0.6 does not accurately predict leak flow when leak qua ity is high.
Experimental data for Test S-07-10B was transmitted 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 discrepancy between the analysis results and the test data.
As a 8
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 3
the CORE 2AL option.
2.
Force the broken loop steam generator secondary pressure to conform to the actual test pressure, since inadequate information is available to predict the secondary side response.
I 3.
Utilize the Homogeneous Equilibrium Model (HEM) discharge model with Cd = 1.0 for two phase break flow.
The post test analysis, described in detail in the following m tion, demonstrates that these model changes substantially improve the predicted results. T
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4.2 Post Test Prediction 4.2.1 Post Test Analysis Model i
The calculations performed and model developed for the pretest prediction were used extensively in the post test analysis.
t1odifications made to the pretest model to obtain the post test model, along with a description of the approach taken, are
'J presented in the following paragraphs.
As shown in Reference 5, the initial conditions for Tests S-07-10B and S-07-100 were essentially the same; thus, the initial conditions used for the pretest prediction were deemed to be valid for the post test prediction. Table 1 contains both l
sets of initial conditions. The ECC parameters, however, 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.
l 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 changes made as required by format changes between version 8.4 and 17.0.
l I
I As noted in Section 4.1, there was uncertainty in the steam generator blowdown calculated in the pretrst analysis.
There were two factors which preclude an accurate prediction of steam l
1 generator secondary blowdown through 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 tha efficiency I
of the steam separator used in the test..
I Lacking this information, it is very difficult to accurately predict steam flow characteristics.
I To compensate for these unknowns, the pressure from the broken h
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.
Since both the intact loop and broken loop steam generators would be forced to follow the same pressure versus time table, heat transfer in the ini.act loop steam generator was stopped at 100 seconds by setting the heat tr/insfer coefficient equal to zero.
This action was.iudged to be acceptable because it had beca deten ined that the primary and secondary systems were somewrat decoupled during the majority of the transient due to high voi1 fractions in the steam generator tubes and resultant poor heat transfer (Reference 5, Page 11).
It was originally noted in the pretest submittal that primary metal heat input appeared to be too large. Since no primary metal temperature histories are available, a detailed evaluation of the primary metal heating concern could not be performed.
However, at 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.
In the pretest
'8 prediction all core heat regardless of flow direction t ;-
deposited in the single core node. When the code predicts
.I substantial core uncovery, as occurred in the S-07-10B analysis, use of the CORE 2AL option produces extremely conservative results relative to core uncovery because the core heat in the uncovered portion of the rods is placed into the remaining.
T.
liquid in the core.
This leads to excessive boiling in the remaining liquid, and thus excessive pressurization.
This assumption is conservative for small break LOCA predictions.
l 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 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 multinode representation, the CRAFT calculated core heat transfer coefficients would now be based upon the inlet fleid condition to the core flow path. Thus, during a core uncovery situation, a low surface heat transfer coefficient would be chosen for the core paths which have steam inlet conditions.
In this ;nanner, the actual heat transfer conditions in the core (i.e. good heat transfer in the covered portion of the rod vis-a-vir, pr>or 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 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 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 HEM model with discharge coefficient equal to 1.0 was used for two-phase fluid and steam conditions.
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1 The resultant post test evaluation noding diagram is illustrated in Figure 2.
Ir suninary, the differences between the post test and pretest models are:
l 1.
The ECC parameters for the post test evaluation were y
changed to represent the actual S-07-100 test condition.
2.
The stean 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 traasfer coefficient was set to zero after 100 ma seconds.
g 3.
A multinode core representation (6 nodes /7 paths) was chosen for the post test evaluation.
4.
The saturated fluid oischarge model was changed from the Moody correlation with a Cd = 0.6 to the HEM correlation wi.h 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 post test analysis, especially as related to core level response.
Figures 3 through 7 present results of the evaluation. Tnese are discussed more fully below.
Figure 3 presents a comparison of system pressure for both the pretest and post test results with the S-07-100 test pressure.
The post test result is very close to the test data.
Especitlly of interest is that the post test curve did not flatten out l
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after 250 seconds like the pretest did. The HPIS came on at 467 l
seconds in the post test prediction as cnmpared to 460 seconds l
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in the test. The high pressure injection system d U not actuate in the pretest prediction.
Examination of Figure 4 shows *.he reason for the differences.
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.
I Figure 5 presents. pressure versus time for the secondary side of the steam generator. As was previously noted, the secondary I
side pressure for both the intact and broken loop was forced to follow actual broken loop test pressure, durine the post test I
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 post 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 l
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i agreement to the measured values than the results provided in l
the pretest prediction. However, the predicted mass flow rate was still somewhat below the actual test data. This is generally a result of the lower system pressure which was predicted as compared to the actual data (Figure 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 predictior, lhis large change is attributed to a combination of the correction in the steam generator blowdown and the multinode core model.
It is important to note that vessel refill, that occurred at *100 seconds due to the loor seal clearing, was calculated to occur in the post test prediction.
Previously, none of the semiscale participants 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 actuation, was quite satisfactory and in good agreement with S-07-10D test data.
Relative to the questions of Reference 2, the following specific conclusions 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 option normally used in the small break LOCA evaluation model needed to be replaced by a more detailed core represt.ntation to obtain best estimate calculations. -
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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 specificaticn of test parameters is necessary before calculations _are performed.
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5.
REFERENCES 1.
Letter from J. H. Taylor (B&W) to Richard P. Denise (NRC), " Analysis Prediction for Test S-07-10B", dated October 9, 1979.
2.
Letter to all 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 L3-1". Doc. No. 51-1125988-00, May 1981.
/
W 4.
D. J. Shimeck, " Analysis of Semiscale M00-3 Small Break Tests S-07-10 and S-07-100", (EG&G Report EGG-SEMI-5201), July 1980.
5.
Letter to J. H. Taylor (B&W Licensic.g) from J. A. Dearier (Code Assessment and Applications Branch NRC) " 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 J. H. Taylor I
(B&W Licensing), " Experiment Data Release - Semiscale Test S-07-10B",
dated March 17, 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 Loss of Coolant, BAW-10092P, Revision 2, Babcock & Wilcox, April 1975.
I 8.
J. R. Paljug. M. D. Charakhani, R. C. Jones, B&W's Best Esti nate Prediction of the LOFT L3-6 Nuclear 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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Table 1.
Initial Conditions and ECC Requirements I
for Semisca: e Test S-07-10D Value Ustd Actual Data Initial Conditions from S-07-10B for S-07-10D Nominal System Pressure, psia 2250 2277 I
Hot Leg Fluid Temperature, F 604.4 605 Cold Leg Fluid Temperature, F
'535.8 541 Core AT, F
68.6 63 l
I Core Inlet Flow, Lbm/sec.
21.4 21 Total Core Power, MJ 1.927 1.94 l
ECC Parameters Intact Loop Accunult. tor (Flood Tank)
B System Pressure at Actuation 232 psia Tank Pressure at Actuation 450 psia 1.6 ft.3 Liquid Volume 3
I Temperature 80 F Cas Volume 0.g8ft.
I Intact T. cop HPIS Actuation Pressure 232 psia Injection Rate (Average) 0.17 Lbm/sec.
Temperature 80 F Intact Loop LPIS Actuation Pressure 305 psia Injection Rate (Average) 0.24 Lbm/sec.
Temperature 80 F
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Table 2.
Sequence of Events i
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Test S-07-10D Pretest Prediction Post Test Prediction l
Event Time (sec.)
Time (sec.)
Time (sec.)
l Blowdown initiated 0
0 l
Pressurizer pressure = 1800 psia 6.5 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 6.0 35-40 Upper plenum liquid 1cvel reaches hot leg 42 25 Pressure suppression system pressure reduction begins 52, 60 Liquid from cold 1 cgs 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,f 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 434 330 467 Lowest point in pest test (93%)
Accumulator injection begins 460 470 llPIS injection begins 460 467 4
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Figurg 7 CORE COLLAPSED LIQUID LEVEL 13.5
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k ESTIMATED UNCERTAINTY IN CORE LIQUID LEVEL 10.0 d
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