ML17268A104

B&W Post-Test Evaluation of Loft Test L3-1.

ML17268A104
Person / Time
Site:	Crystal River
Issue date:	05/31/1981
From:	Rosalyn Jones, Savani N BABCOCK & WILCOX CO.
To:
Shared Package
ML17268A103	List: ML17268A102 ML17268A104
References
TASK-2.K.3.30, TASK-TM 51-1125988-01, 51-1125988-1, TAC-45817, NUDOCS 8106160506
Download: ML17268A104 (21)

Text

BEW's Post Test Evaluation of LOFT Test L3-1 Document No. 51-1125988-01 May 1981 Principal Investigators N. K. Savani R. C. Jones Prepared by BABCOCK & WILCOX COMPANY for The Owners Group of Babcock E Wilcox 177 and 205 FA NSS Systems

'8106160

$ ~g

TABLE OF CONTENTS

~Pa e

1. I NTROOUCTION

2.

SUMMARY

2

3. OISCUSSION OF RESULTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3

4. CONCLUSIONS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 10

5. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . 12 6o TABLE o ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 13

7. FIGURES ..... '.................... 14

l. INTRODUCTION B&W's pretest predictions for LOFT Test L3-1 was submitted to the NRC by Reference 1. The test was designed to provide data for a PWR simulated 2.5 percent single ended break in the cold leg with ECCS injection limited to the intact loop. The break was large enough to cause system depressurization.

The test has been. completed and the comparisons of pretest predictions to the test data have been published in Reference 2. Inspection of this comparison shows that certain discrepancies exist between the pretest predictions and the test data. Based on these discrepancies, Mr. R. W. Reid of the NRC sent a letter to all Babcock E Wilcox Licensees (Reference 3) requesting the submission of a post test evaluation of LOFT Test L3-1 and Semiscale Test S-07-10D. According to the letter, the objectives of this post test evaluation are:

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 predict the test data,

3. Assess whether any improvements and/or modifications necessary for 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 improvements to the test facility, instrumentation, or test procedures to improve the verification process.

This report is in response to Mr. Reid's letter. It provides MW's post test evaluation of LOFT Test L3-1 and addresses the above mentioned objectives. The post test evaluation of S-07-10D is provided in Reference 4.

2.

SUMMARY

This report is a partial response to Mr. R. W. Reid's letter requesting the submission of a post test evaluation of LOFT Test L3-1 and Semiscale Test S-07-10D. This report provides the post test evaluation of the LOFT 3-1 Test.

LOFT Test L3-1 was a 2.5 percent single ended cold leg break with a nuclear core initially operated at a power of 50 MW. The pretest prediction of the test was performed using the CRAFT2 computer code and was submitted to the NRC by Reference l.

The pretest prediction performed by HEW compares favorably with the test data. However, it is recognized that the comparison shows that cetain discrepancies exist between the pretest predictions and the test data. A detailed review of the results of L3-1 is performed and the sources of these discrepancies are identified. gualitative justification is provided in this report demonstrating that if these sources of discrepancies were eliminated, the prediction would be in better agreement with the test data.

Relative to the objectives of the post test evaluation identified in Reference 3, the post test evaluation of LOFT Test L3-1 confirmed that the CRAFT2 computer code can predict the small break LOCA phenomena observed in the test, provided that adequate test conditions are provided. No code modifications are believed to be necessary for the prediction. 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 "blind pretest" model, release of the test data, and then a post test evaluation where the changes from the "blind pretest" model must be justified.

3. DISCUSSION OF RESULTS 3.1 Pretest Prediction Via Reference 1, B&M's pretest prediction of LOFT Test L3-1 was submitted to the NRC. A detaile'd comparison of the prediction to the actual data was performed by EG&G and is documented in Reference 2. As is shown by that report, the B&W prediction compares favorably to the experimental data.

LOFT Test L3-1 was a 2.5 percent single ended cold leg break with a nuclear core intially operated at a power of 50 NM. The major components of the LOFT facility are shown in Figure 1. The CRAFT2 noding diagram utilized for the L3-1 pretest prediction is shown in Figure 2 and described in Table l.

The LOFT L3-1 system description and the initial conditions used for the pretest prediction were obtained from References 5- and 6. Ouring the L3-1 test, some of the actual conditions were different than the initial conditions provided by EG&G before the test. These differences between the initial conditions and the actual conditions are described below:

1. In the test, the auxiliary feed pump was started at 75.0 seconds while in the B&M pretest prediction, the .auxiliary feedwater was started at 60.0 seconds based on the initial conditions provided by EG&G.

2. The specified initial accumulator pressure in Reference 6 was 600 psig. The actual measured pressure of the accumulator during the test was 633 psig.

3. During the test, two leakage paths occurred, which were not identified to the participants who were attempting to predict the test. 'ence, these leakage flow paths were not modeled in the BKW's pretest prediction model.

4. No specific information about the steam flow control valve was provided in the initial conditions, e.g. flow area, the percent initially open, how fast \ the valve closed. No specification on leakage through the steam flow control valve was provided. Examination of more recent LOFT Tests, L3-7 and L3-6, indicate that there is some leakage through the main steam control valve when it is in its fully closed position (Reference 7) .

Based on conversations with EGIIG personnel, the steam valve was assumed to be 75 percent open, and to close at a rite of 5 percent per second to the fully closed position. The effective steady-state flow area was calculated based on the Moody Critical Flow Model and area was changed as a function of secondary side pressure. No leakage through the valve was allowed when the valve was in its fully closed position.

As can be seen, there are several significant discrepancies between the initial conditions assumed in the analysis and the actual system conditions of the experiment. The impact of these differences on the prediction is .

described in the following section.

3.2 Post Test Evaluation In this section, the comparison of 8&M's pretest prediction with the test data is provided. Along with this comparison, the discrepancies between 8&M's prediction and the test data are explained based upon the differences in assumed and actual system conditions described above.

3.2.1 Steam Generator Secondar Pressure In this test, the steam gener ator secondary side pressure quickly rose after the reactor scram due to'he closing of the steam flow control valve. In the test this valve did not open during the initial 1500 seconds of the transient. The auxiliary feed pump was started at 75 seconds cooling the mixture in the steam generator.

The steam generator secondary side pressure comparison of the pretest prediction and the test data is shown in Figure 3. It is obvious that the calculated secondary pressure exhibited the same trend as the test data; however, the pressure was underpredicted slightly for the first 250 seconds and then overpredicted from 400 seconds on.

As explained earlier, in B&M's pretest prediction, the steam flow control valve was assumed to be 75 percent open and to close at a rate of 5 percent per second to the fully closed position. The auxiliary feedwater was started at 60 seconds.

In the actual test, the auxiliary feedwater was started, at 75 seconds. No specific information about the steam flow control valve performance is provided to compare it with the B&M's pretest prediction model assumption.

By modeling the steam flow control valve exactly as it behaved in the actual test, the steam generator pressure prediction is expected to improve. It is felt that the steam flow control valve was probably open less than 75 percent initially and thus, was fully closed earlier than calculated. The initiation of auxiliary feedwater at 75 seconds, as it happened in the test, is expected to bring the predictions in better agreement with the test data during the early part of the transient because the steam generator cooling will be delayed.

Examination of recent LOFT tests indicate that there is leakage through the steam flow control valve when it was in its fully closed position (Reference 7). In the BEW pretest prediction model, no leakage through steam flow control valve was allowed. It is expected that there was probably leakage in the L3-I test. Inclusion of steam flow control valve leakage would result in a faster steam'enerator depressurization more typical of the experimental data.

3.2.2 Intact Loop Accumulator Pressure The accumulator injection during the test was initiated at 634 seconds after the rupture, The measured initial pressure was 633 psig. The specified

-initial accumulator pressure in Reference 6 was 600 psig.

Figure 4 shows the comparison of the B&W's calculated accumulator pressure to the test data. As can be seen, BKW's predicted accumulator pressure response shows a trend similar to the test data indicating that the accumulator emptied at a rate similar to the experiment. The discrepancy prior to the initiation of the accumulator is due to the difference between the initial conditions given and the actual test conditions. BEW's pretest prediction resulted in the initiation of the accumulator flow at 850 seconds. The initiation of accumulator flow just due to the proper initial accumulator pressure alone will result in ~100 second earlier initiation of accumulator flow. Also, the expected improvement in the system pressure prediction by utilizinq the actual test conditions in the pretest prediction model (as discussed in Section 3.2.4) will further improve the predicted t

time of accumulator actuation.

3.2.3 Break Flow Figure 5 compares the leak flow rate of BEW's pretest prediction and the test data. The predicted break flow exhibits the trend of the data for the initial 50 seconds and from ~ 500 seconds on. The pretest prediction over these time frames is within the range of the test data. The predicted break flow between 50 to 250 seconds is essentially constant and significantly higher than the test data. The predicted break flow then sharply drops and

~

is slightly lower than the test data up to 500 seconds.

This discrepancy between the predicted break flow and the test data is primarily due to the leakage paths that existed in the test. Ouring the test, leakage through the cold leg warm-up line, reflood assist bypass valves and from hot leg to downcomer occurred. These0 leakage paths were not

~

identified to the participants who were attempting to model the test.

Because of these leakage paths, there was a direct path for the hotter fluid of the system to be transported and mixed with the fluid at the break. ~

Consequently, subcooled liquid discharge will end earlier and there will be a smoother transition through two phase flow to steam, which was typical of the actual experiment data.

In BEW's pretest prediction model. these leakage flow paths were not modeled. Therefore, the predicted break flow during the first 250 seconds of the transient is subcooled liquid's no steam relief path to the break is available until the loop seal is. cleared. The clearing of the loop seal will decrease the break flow sharply as a result of a sudden decrease in the density of the fluid at the break.'he slightly underpredicted break flow between 250 to 500 seconds is due to the discharge model utilized in the prediction model and as a result of the earlier difference between the prediction and the test data. The Bernoulli-Moody Discharge Model with a discharge multiplier of 0.6 for subcooled and two-phase flow was utilized in the prediction. The discharge multiplier was changed to 0.9 when the break flow stabilized at pure steam. In order to predict the two-phase discharge flow precisely, the discharge multiplier should be changed as a function of break fluid quality. The expected transient of the test without the leakage flow paths was a long duration subcooled liquid discharge followed very quickly by a transition to steam. Hence the two-phase discharge was expected to occur only for a short time during the transient. Based on this, the discharge multiplier for the entire two-phase discharge flow was kept constant at 0.6. The discrepancy between the predicted break flow and the test data during 250 to 500 seconds is essentially due to the constant discharge multiplier of 0.6 for the entire two-phase flow and the earlier difference between the prediction and the test data.

In short, the predicted. break flow compares reasonably well with the test data. The prediction is expected to improve significantly if the actual test conditions (leakage paths) and the more appropriate discharge model for the test are utilized in the prediction model.

3.2.4 S stem Pressure In the test, the system pressure was characterized by a rapid depressuriza-tion to the system fluid saturation pressure. The. depressurization rate then decreased until 300 seconds when the system liquid level decreased below the break orifice, allowing steam to escape and the system to depressur ize more rapidly. From 1000-15000 seconds depressurization rate was decreased as a result of the liquid level in the broken loop increasing above the break orifice.

Figure 6 shows the comparison of BEW's pretest prediction with the test data. The prediction generally shows the same trend as the test data and is in good agreement with the test data. However, it is recognized that the pressure is slightly under predicted from 50 to 200 seconds and over predicted from 600 to 1100 seconds.

The initial underprediction of the pressure is mainly due to the fact that steam generator secondary side pressure was underpredicted during this time frame. During this period of time, the secondary side pressure controlled the primary side pressure. Thus, the expected improvement that could be obtained in the steam generator pressure would imprbve the primary system pressure comparison. The long term overprediction of system pressure is due to a combination of various factors. As mentioned earlier, the pretest prediction did not include the leakage paths. With the addition of these leakage paths, as explained in Section 3.2.3, the break flow prediction can be improved and this would improve the system pressure with the test data.

As discussed in Section 3.2.2, BEW's pretest prediction of accumulator initiation was later than the acutal test. This was partially a result of the difference between the initial condition given and the acutal test condition of the accumulator. With the initiation of the accumulator, steam will be condensed and hence a more rapid system depressurization will take place. The delayed predicted accumulator actuation contributed to the over prediction of the system pressure.

The steam generator secondary side became a heat source in BEW's pretest prediction at about 400 seconds. During the pretest prediction, no leakage through the steam flow control valve was allowed once it was in its fully closed position. If in the actual test, leakage throuqh the steam valve existed once it was in its fully closed position, a. portion of the secondary side energy would be released through the leakage and hence less energy would be added to the primary system. This also contributes to the over-prediction of the system pr essure in 8&M's pretest prediction model.

In short, BEM's pretest prediction of the system pressure compares reasonably well with the test data. The difference between the prediction and the test data is a result of the differences between the assumed and actual test conditions.

4. CONCLUSIONS In general, BEM's pretest predictions of the L3-1 test are in good. agreement with the test data. Review of the comparisons of, BEM's pretest prediction and the test data by EGEG in Reference 2, reveals that the predictions exhibited the same trend as the data for all the parameters compared. It is recognized that the comparison shows that certain d'ihcrepancies exist between the predictions and the test data. The differences between the predictions.and the data are due to differences in the assumed and actual system conditions. The qualitative analysis provided in this report demonstrates that incorporation of the actual'xperimental conditions would result in an improved prediction.

Relative to the objectives of the post test evaluation, as stated in the

=- February 24, 1981 Reid letter (Reference 3), the following conclusions are drawn:

The evaluation of LOFT Test L3-1 confirmed that the CRAFT2 computer code is capable of predicting the small break LOCA phenomena observed in the test if the actual test conditions were utilized.

.2. No code modifications are believed to be necessary to predict the test.

3. Since no code modifications are necessary to predict the test, no modifications to the present ECCS small break evaluation model have been identified from this experiment.

5. REFERENCES

1. Letter from J. H. Taylor to D. F. Ross, "B&W's Best Estimate Prediction of the LOFT L3-1 Nuclear Small Break Test Using the CRAFT2 Computer Code," December 13, 1979.

2. Czapary, L. S., "LOFT L3-1 Preliminary Comparison Report" EGG-CAAP-5255, dated September 1980.

3. Letter to All Babcock & Wilcox Licensees from R. W. Reid, Chief Operating Reactor Branch 84, Division of Licensing, February 24, 1981.

4. T. E. Geer, et.al., B&W's Post Test Analysis For Semiscale Test S-07-100 Doc. No. 86-1125888-00, May 20, 1981.

5. LOFT System and Test Description (5.5 ft. Nuclear Core 1 LOCES),

NUREG/CR=0274, TREE-1208, Reeder, D. L. dated July 1978.

6. Information for L3-1 Standard Problem, Kaufman, N. C., Kau-203-79, dated September 28, 1979.

7. Modro, S. M. & Condil, K. G., "Best Estimate Prediction for LOFT Nuclear Experiment L3-5/L3-5A," EGG-LOFT-5240, dated September 1980.

Table 1. Node Descri tion Node No. Description 1 Downcomer annulus 2 Downcomer 3 Lower plenum 4 Core 5 Upper plenum 8 upper head 6 Pressurizer 7 Hot leg, intact loop 8 Steam generator, front half of primary side, intact loop 9 Steam generator, back half of primary side, intact loop 10 Steam generator outlet, cold leg piping, intact loop 11 Pump suction, intact loop 12 Pump discharge,'intact loop 13 Hot leg plus half simulated SG, broken loop 14 Half simulated SG plus half pump, broken loop 15 Half pump plus piping, broken loop 16 Leak node 17 Suppression tank 18 Secondary side, intact loop

~ ~

intact loop 8roken loop Quick opening valve {2)

Steam generator simulator Pressurizer Steam generator Isolation valve

~

{2)

~

J

~ \

r r~

4

~4 Pump r'~

FCC injection simulator Iocalion Pumps Reactor vessel Dov.ncomer Suppression vessel Core Lower plenum

~ M1 Reaclor vessel >>NEl. blOOOf4 Figure I. IA)t i moJoroomponents.

TO NODE 17 Q]1 Q Q Qzo 18 QO Q

Q1O Q

14 13 Q18 17 1S 10 Q4 12 I

Q" I

1O 25 Q4 QO 30 Q Q1O QEO QOO QE QO1

Q LP1 21 ACC Q Q 3 22 23 17 MOO E P16 lEAK OOOO, FLOW PATHS 32 8 33 ARE LEAK FLOW PATHS LOFT L3-1 NODING DIAGRAM Figure 2

1250 PT-PW IOA UPPER BAND LOWER BAHO BKW 1000 pe% ~

750 4

500 0 500 1000 1500 Time After Rupture (Sec)

COMPARISON OF CALCULATIONS OF STEAM GENERATOR SECONDARY SIDE PRESSURE TO DATA Figure 3

750 PT-P I 20-Q3 UPPER BAND LONER BAND lQ 500 D

8 0

250 0 500 1000 1500 Time After Rupture (Sec)

COMPARISON OF B8W CALCULATIONS OF INTACT LOOP ACCUMULATOR PRESSURE TO DATA Figure 4

TTE<<BL- I UPPER BAND LONER BAND 40 0

30 3

0 20 K

10

~ Co ~ 5 0

0 500 1000 1500 Time After Rupture (Sec)

COMPARISON OF B8W,CALCULATIONS OF BREAK MASS FLOW TO OATA Figure 5

2000 PE- I UP- I UPPER BAND LOWER BAND BKW

- 1500

~ 1000 500 0

0 500 1000 1500 Time After Rupture (Sec)

COMPARISON OF CAlCULATIONS OF UPPER PLENUM PRESSURE TO OATA Figure 6