Text

Best Estimate Post-Test Prediction for Loft Nuclear Experiment L3-6
	ML20030B499
Person / Time
Site:	Yankee Rowe, Maine Yankee
Issue date:	08/31/1981
From:	Harvey R, Husain A, Schor L; YANKEE ATOMIC ELECTRIC CO.
To:	;
Shared Package
ML20030B496	List: ML20030B495; ML20030B499;
References
	YAEC-1273, NUDOCS 8108180209
	Download: ML20030B499 (62)
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BEST ESTIMATE POST 'IEST PREDICTION FOR LOFT NUCLEAR EXPERIMENT L3-6 by Liliane Schor i

Robert C. Harvey Ausaf Husain August, 1981 1

D b I 9 91 Pr e pa. et' By 2

Liliane Schor Date LOCA Group, Nuclear Engineering

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I Prepared By

Dat'e Robert C. Harvey

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LOCA Group, Nuclear B4gineering Prepared By Wca_b h ert.

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Ausaf Husdin, Manager

'Date LOCA Group, Nuclear Engineering Reviewed By j jsso % sh,

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Tom Fernandez

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LOCA Croup, Nuclear Engineering

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Approved By Bruce Slif er, Manger'

' Ddte Nuclear EngineeriQ, Yankee Atcznic Electric Company Nuclear Services Divisien 1671 Worcester Road I

Fr amingh am Massachusetts 01701 I

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DISC 1 AIMER OF RESPONSIBILITY This document was prepared by Yankee Atomic Electrie Company for its own use.

It is being made available to others as a public service without monetary or other compensation to Yankee, upon the express understanding that neither Yankee Atomic Electric Company or any of its officers, directors, agents or empicyees assumes any obligation, responsibility or liability, or I

meikes any carranty or representation, with respect to the contents of this document or its accuracy or completeness.

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I ABSTRACT I

This rep.rt contains Yankee Atomic Electric Company's best-estimate p.st test analyses of LOIT test L3-6.

The results of these at%1yses dem.nstrate the adequacy of YAEC's

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ACKNOWLE DGEMENTS i

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The authors wish to express their sincere thanks to Dr. Tom Fernandez for reviewing this report and for his many helpful comments and suggestious thr ou gh out the completion of this work. Thanks are also due to Ms. Susan l

l Henchey for typing this report.

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I TABLE OF CONTENTS i

i Page I

DISCIAIMER OF RESPONSIBILITY..............................

ii Ar3 TRACT..................................................

iii ACKN0WIIDGEMENTS..........................................

iv TA B LE O F C 0NTE NT S.........................................

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4 LI ST OF TA B LE S............................................

vi LI ST O F F IG U RE S........,..................................

vii 1.0 I N TRO D UC T I O N..............................................

1

2.0 DESCRIPTION

OF L3-6 M0 DEL.................................

3 2.1 Computer Program...................................

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2.2 LOFT Model Setup...................................

3 2.2.1 P r ima ry Sy s t em Mo d eli ng............................

4 2.2.2 Sec ondary Sys tem Mod eling..........................

6 2.2.3 Break Flow Modeling................................

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3.0 RE SULT S OF TH E P OST-IE ST ANALYSI S.........................

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3.1 P r ima ry S y s t em P r e s su r e..............................

10 3.2 Secondary System Pressure............................

10 3.3 Break F1ow...........................................

10 g

3.4 Primary Coolant System Mass Balance..................

11 g

3.5 Densities............................................

11 3.6 Differential Pressures...............................

12 3.7 Fluid Temperatures...................................

12

4.0 CONCLUSION

S...............................................

13 5.0 R E FE R E NC E S................................................

14 6.0 APPENDIX - INPUT LISTING..................................

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LIST OF TABLES i

Ntznber T i' le_

Page Table 1 RELAP4/ MOD 3 LOFT System Description 15 Table 2 Predicted Primary Sytem Coolant Mass Balance 17 at 2000 seconds I

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I LIST OF FIGt;RES i

Ntsnber Title Page Figure 1 RELAP4/ Mod 3 LOFI Nodalization 18 Figure 2 Primary System Pressure (0.65 Moody) 19 Figure 3 Density in Intact Loop Cold Leg (0.65 Moody) 20 Figure 4 Break Flow Rate (0.65 Moody) 21 Figure 5 Primary System Inventory (0.65 Moody) 22 Figure 6 Comparison of Primary System Pressure with 23 and without Quality Enhancement Figure 7 Conparison of Break Flow Rate with and without 24 I

Quality Enhancement Figure 8 Comparison, of Primary Systeo Pressure; 25 0.65 Moody and 0.95 HEM Figure 9 Comparison of Break Flow Rate; 26 0.65 Moody and 0.95 HEM Figure 10 Primary System Pressure 27 Figure 11 Secondary System Pressure 28 Figure 12 Break Flow Rate 29 i

Figure 13 Primary System Mass Inventory 30 Figure 14 Density in Intact Loop Hot Leg 31 Figure 15 Density at Steam Generator Exit 32 Figure 16 Density in Intact Loop Cold Leg 33 Figure 17 Differential Pressure in Intact Loop Across 34 I

Reactor Vessel Figure 18 Differential Pressure in Intact Loop Across 35 Primary Coolant Pumps Figure 19 Differential Pressure in Intact Loop Across 36 Steam Generator

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Figure 20 Coolant Temperature in Intact Loop Hot Leg 37 I

Figure 21 Coolant Temperature in Intact Loop Primary 38 Side Inlet and Outlet Plenums

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1.0 INTRODUCTION

LOFT Test L3-6, performed on December 10, 1980, simulated a four inch diameter break at the pump discharge in the cold leg of a commercial PWR.

I The United States Nuclear Regulatory Commission (NRC), in Reference 1, required that Yankee Atomic Electric Company (YAEC) submit a " blind" analysis o f LOFT te s t L3-6.

In order to ascertain that the analysis was a truly blind prediction, the NRC required the submittal of the model to be used prior to I

the L3-6 test date.

YAE C me t this pre-test requirement on December 5,1980 in its submittal to the NRC of the L3-6 model (Reference 2).

On January 15,1981, EG6G Idaho britied the NRC and the holders of licensed ECCS models on the red.cs of LOFT Test L3-5 (purrys tripped) and L3-6 (pumps running). After ttis meeting the L3-6 Quick Look Report (Reference 3) and the Experiment Data Report (Reference 4) were released. The test changed from a " blind" analysis to a " blind post-test" analysis in which minimal I

changes, such as incorporation of the actual initial conditions and modification of the break discharge coefficient, were allowed.

On March 3,1981 YAEC submitted to the NRC the " blind" post test prediction of LOFT test L3-6 (Reference 5).

On May 18, 1981, the NRC invited Yankee to give a presentation on the LOFT test L3-6 prediction. Due to discrepancies between measured and predicted primary system pressures, the NRC requested YAEC to reanalyze the test and to make another analysis submittal.

Augus u 1,1981 was established as the resubmittal date.

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The RELAP4/ MOD 3 "Dlind" ; ost-test prediction [5] and experimental data from LOFT test L3-6 were used to identify changes required to the blind analysis model in order to improve the code predictions. This report presents the post-test analysis results and discer.ssion of some of the sensitivity studies, which were conducted to arrive at the final model.

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2.0 DESCRIPTION

OF L3-6 MODEL This section describes the new LOFI model used by YAEC to predict the L3-6 test.

2.1 Computer Program The thermal hydraulic analysis of L3-6 was performed using a modified version of Yankee Atomic Electric Company's RELAP4/ MOD 3 licensed code [6].

The LOFT prediction specific modifications, made to the licensed code, were I

documented in Reference S.

2.2 LOFT Model Setup I

The analytied model used in the post-test prediction of LOFT L3-6 is based on the " blind pos t-test analysis submitted on March 3, 1981 [5].

Modifications were made to the " blind" post test model to reflect actual test conditions and peculiarities specific to the LOFT facility. These changes are described in the following subsections.

I A nodal diagram for the post-test model is shown in Figure 1.

A description of the volumes, junctions and heat slabs is given in Table 1.

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I 2.2.1 Primary System Modeling I

Primary system pressure comparisons between the " blind" post test analysis (5) and experimental data showed reasonable agreement up to 900 seconds, after which the analysis predicted faster depressurization than the data.

This f aster depressurization was caused by an. brupt transition in break flow quality from a low to high value. This transition occurred as a result of the uncovering of the hot leg nozzles. Although in the experiment, the break flow did ultimately go to tigh quality flow, the transition occurred much later and was less abrupt. There was also an anomalous behavior in pressure at about 1300 seconds whereby RELAP4 predicted a 30 psia repressurization of the primary system.

I The reason for these differences between data and analysic was due primarily 70 the Wilson bubble rise rodel used in the vessel volumes which predicte/ core uncovery. To remedy these deficiencies in the LOFT prediction, an ana.ysis of flow patterns and slip ratios in the vessel volumes was performed. Migh slip ratios are estimated in the upper head and lower plenum regions while a more homogeneous mixture is predicted in the core. Since RELAP4/ MOD 3 dees not have a slip option, a Wilson bubble rise model was used in the lower plenum and upper head volumes. All other vessel and primary system pipe. volumes were treated as homogeneous components.

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Additional changes to the primary system representation were made to better reflect the LOFT system and are as follows:

5.

Revised vessel dimensions The dimensions of the LOFT reactor vessel were revised based on the new vessel dimensions provided by EC&G as a modification to Reference 7.

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Upper head bypass A flow path was added between the upper head and the downcomer.

The area was adjusted to allow for 3% bypass of the initial primary system flow rate.

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Wall heat I

The effect of wall heat was introduced into the primary system by modeling all the pipe metal in the intact and broken loop. The wall heat in the reactor vessel was re-ised to account for the downcomer filler material.

l (ur 4.

Two phase pump head and torque degradation The YAEC " blind" post test analysis of LOFT test L3-6 used the two phase pump head and torque degradation multipliers which were baseo on the CE/EPRI punp performance tests (Reference 8).

A new set of head and torque multipliers have been developed by EC&G, I

Idaho (Reference 9).

These multipliers were obtained from curves of I I

I the LOFT pump head and h draulic t.oroue ratios (two phase head and torque with respect to their ringle-phase liquid values) assuming the " fully degraded" two-phase head and torque are zero.

A comparison of the two sets of curves was made and it was concluded that the EG&G curves are more appropriate for the LOFT L3-6 analysis.

I-2.2.2 Secondary System Modeling I

In the "bliM" post test prediction, the steam generator secondary boiler section was treated as one volume using the Wilson bubble rise model. This I

modeling techni que gave good agreement with data during the first 900 sec.

Therea f ter, the system was predicted to be in reverse heat transfer mode, (secondary pressure exceeded primary pressure) and the predicted and measured secondary side pressures diveg

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I Moreo ver, the " blind" post-test prediction has shown that past 900 seconds, the auxiliary feedwater was predicted to flow from the downcomer into the steam dome, bypassing the boiler section. This c w:. 4 condinsation of steam in the steam dome which in turn caused excessive depressurization in the l

boiler region. This unphysical behavior was eliminated by allowing only forward flow in the junction connecting the steam dome to the downcomer.

I LOFI test L3-6 data shows temperature stratification in the steam I

generator secondary, beginning at about 400 seconds and becoming more l

l pronounced with time. These temperature gtadients are due to the injection of i

cold auxiliary f eedwater, which collects at the bottom of the steam generator l

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I and forms a subcoM.ed layer. Atop this subcooled layer, there is water at a saturation temperature corresponding to the secondary system pressure and steam in steam dome.

Forvard heat transfer (primary to secondary) takes place in the lower s abcooled layer and reverse heat transfer occurs above it.

The measurements of prir.ary fluid temperatures in the steam generator inlet and outlet plena show there is very little net heat transfer to the primary system during this period of forward and reverse heat transfer. As the auxiliary f eedwater collects in the bottom of the s team generator secondary, the level rises causing a compression eff ect.

This compression tends to decrease the rate of depressurization produced by condensation in the steam dome. To be able to simulate this course of events, a steam generator secondary with five I

volumes in the boiler region was utilized.

I Changes in the modeling of the steam generator secondary:

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Nodalization changes I

The boiler section was represented wth five vertically stacked volumes. The Wilson bubble rise model was used in modeling the upper most volume. The junction connecting the steam dome to the downcomer was modeled such as no reverse flow was allowed.

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Steam relie f valve leakage Instead of a leak table versus time, the post-test analysis allows the code to calculate its own leak rate through the team relie f valve. Thi as accotrplished by imposing the desired experimentally measured pressure in a volume connected to the steam I I

I dome. The junction between the two volumes simulated he leak through the relief valve. Re flow in this junction was not a', owed to reverse since this would have represented mass addition to the steam generator secondary.

3.

Wall heat The eff ect of wall heat was introduced into the steam generator

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secondary side nodes by it. luding the steam generator vessel wall as heat slabs.

I 2.2.3 Break Flow Modeling I

An initial attempt was made to simulate the break flow rate using the Henry-Fauske model during subcooled blowdown and the Moody model during two phase critical flow. Contract ion coefficients of 1.0 and 0.65 were used, respectively.

The calculations using the Moody model did an outstanding job I

of predicting the pressures and censities in the primary system (Figure 2 &

3).

However, the break flow rate was overpredicted for the first 750 l

seconds, and was slightly underpredicted later on in the transient (Figure 4).

To improve the break flow, a break quality enhancement model was I

implemented in RELAP4/ MOD 3.

The quality enhancement function was based on a ph e nome no n, demonstrated by the test data, in which the quality in the break l

spool piece, upstream of the break orifice, was higher than the quality in the cold leg. nis was a result of unequal velocities of the steam and water droplets as the two phase mixture executed a 90 turn in order to enter the i

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break spool piece.

The quality enhancement function was deriveJ from measured 1

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I densities in the cold leg and spool piece. The adjustment of the break quality with respect to the cold leg quality had no effect on the primary system pressure while it Improved the break flow r.ite by about 0.5 lb/sec at 800 seconds. This sensitivity study which was performed for the fi-c 900 seconds, is shown in Figures 6 and 7.

I A further attempt to improve the break flow was made by replacing the Moody model with the hcmogeneous equilibrium model (HEM). A contraction coefficient of 0.95 was used in conjunction with the HEM model. Comparisons between the primary system pressure and the break flow using the two critical flow models are provided in Figures 8 and 9, respectively.

I Since the primary system pressure was not greatly atfected by the HEM model, and the predicted break f bw is closer to the measured data, the HEM model was used throughout the analysis. Additionally, the use of the Moody model resulted in lower moss inventory le f t in the system as compared to the measured di '.?

(Figure 5).

This result is a direct consequence of the higher break flow rate calculated by the Moody model as described above.

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I 3.0 RESULTS OF THE POST-TEST ANALYSIS This section presents comparisons between the experimental data provided by EG6G (Reference 10) ar.d the " post-test" prediction. The selection of the parameters provided in this report was based on the availability and quality of the data, the correspondence between modeled volumes and physical locations of measurements, and the overall importance of the parameter in assessing the capability of our small break methodology.

I 3.1 Primary Ccolant System Pressure Figure 10 compares the predicted masured primary coolant system pressure. Bere is good agreement between the analysis and the data for most 1

of the transient. A maximum variance of 25 psi between the measured and the predicted pressure is seen at about 500 seconds which is well within the estimated pressure error bands of +39 psia.

3.2 Secondary Side Pressure I

Figure 11 compares the predicted and measured pressures in the secondary system. The agreement of analysis with data is excellent through out the transient.

I 3.3 Break Flow I

The comparison between calculated versus measured break flow is presented in Figure 12.

The analysis overpredicts the break flow up to about 750 seconds, af ter which the calculated break flow follows the data rather well. I

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3.4 Primary Coolant System Mass Balance LOFT test L3-6 demonstrated the maintenance of adequate core cooling during a loss of coolant event in which the pumps were powered, even when a major portion of the primary fluid had been lost. Af ter the pumps are t r ipp ed, the ability of the system to provide core cooling depends, in part, on the amount of mass left in the primary coolant system and the rate at which ECC water can be injected into the core region. The L3-6 post test analysis predicted a mass inventory of 1347 lbm at 2000 seconds. This appears to be in good agreement with the final inventory of 1430 lbm (+140 lbm) reported by EG&G in Reference 3.

Figure 13 presents the comparison between calculated and measured mass inventory in the primary coolant system. Table 2 summarizes the mass balances of the post-test analysis prediction.

I 3.5 Dens it ie s I

Densities in intact loop cold and hot legs are presented in Figurec 14 through 16.

Thr ough out the trans ient, densitites are slightly overpredicted.

It is to be noted that in the sensitivity studies where the Moody critical flow model was employed, an excellent prediction of densities throughout the loop was achieved (Figure 3).

With the HEM model, we achieved better agreement on br.eak flow at the expense of higher densities in the primary c oolan t system. To achieve agreement on both the parameters ( break flow and density) a two velocity model which would increase the quality of the primary fluid leaving the reactor vessel, may have been needed. This phenomenon is I

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I confirmed by analysis of the test data at the exit of the steam generator and loop seal where relative velocity ratios (V -V /V ) of 3% to 14% were g g c alc ula t ed.

Moreover, slip ratios of 1% to 15% were calculated in the reactor vessel using the RELAP4/ MOD 3 preliminary results. However, the inclusion of a slip option ti. RELAP4/ MOD 3 was beyond the scope of this analysis.

I 3.6 Differential Pressures.

I Differential pressures across core, pump and steam generator intact loop are presented in Figures 17 through 19.

RELAP4/ MOD 3 follows the trends of the exp er ime n t, yet it overpredicted the data until about 1100 seconds.

I 3.7 Fluid Temperatures Comparison between predicted and measured temperatures in intact loop hot leg and in the primary side steam generator inlet and outlet plenums are presented in Figures 20 and 21.

1 RELAP/+/ MOD 3 predicts the LOFT data well.

The discrepancies observed past 1250 seconds are thought to be caused by pressure and temperature instrumentation errors or possible radiation from the hot walls to the thermoc ou ple s. The difference of 15 F at 2000 seconds would correspond to a difference in pressure of 70 psia which contradicts the accurate prediction of the primary system pressure at 2000 seconds (Figure 2).

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4.0 CONCLUSION

S I

The post-test analysis presented in Section 3 was performed using information obtained frcan analysis of the L3-6 data. This enabled us to study the behavice of the LOFT system when the pumps are running. The post test analysis did an outstanding job in predicting the primary coolant system pressure and the secondary system pressure. The break flow is slightly overpredicted for the first 750 seconds while the agreement was excellent thr ou ghou t the rest of the transient. This good prediction of break flow gives calculated primary coolant system mass inventory well seithin the error band of the data.

The resui ts of these analyses demonstrate the adequacy of YAEC's methodology to predict small break LOCA experiments and these methods can be successfully applied in the analysis of small break loss of coolant accidents in large PWR's.

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I 5.0 REFE RENCES 1.

USNRC letter to YAEC, dated June 26, 1980,

Subject:

Prediction Requirement of LOFT Small Break Test L3-6 2.

YAEC letter to USNRC, dated December 5,1980,

Subject:

Input for LOFT Small Break Test L3-6 Prediction I

3.

Glenn E. McCreery, " Quick Look Report on LOFT Nuclear Experiments L3-6/L8-1", ECC-LOFT-5318, December 1980.

I 4.

P. D. Bayless, J. M. Carpenter, " Experiment Data Report for LOFT Nuclear Small Break Experiment L3-5 and Severe Core Transient Experimen L8-1",

NU REG /C R-1868, Janua ry, 1981.

I 5.

L. Schor, J. Loomis, A. Husain, "Best estimate Post Test Prediction for LOFT Nuclar Experiment L3-6", March 20,1981.

6.

Yankee Atomic Electric Company WREM based PWR ECCS Evaluation Model (version YAEC-05B), YAEC 1160, July 1976.

7.

D. L. Reeder, " LOFT System and Test Description (5.5 f t Nuclear Core 1 LOCES)", TEE-1208, July, 1978.

I 8.

EG6G presentation to NRC on January 15,1981 (RELAP5 Calculations of L3-5 and L3-6).

j 9.

T. H. Chen, " Primary Coolant Pump Performance During LOFT L3-6 5

E xpe r imen t",

March 2, 1981.

10.

LOFT L3-6 Data Tapes.

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Table 1.

RELAP4 Blowdown System Model Description Control Volume Description 1

Nuclear Core 2

Core Bypass 3

Steam generator secondary steam dome 4&5 Upper Plenum 5&7 Intact loop hot leg I

8 & 13 Steam generator inlet plenum and outlet plenum 9 & 12 Straight section of steam generator tubes 10 & 11 Curved sections of steam generator tubes 14 Steam generator outlet piping 15 Piping leading to the tee p eceding the coolant pumps.

16 Piping from tee to primary coolant pumps 17 Primary coolrnt pumps 18 & 19 Intact loop cold leg 20 Upper annulus at the vessel inlet I

21 D ownc omer 22 Lower plenum 2? & 24 Broken loop cold leg 25,26,27 & 28 Broken loop hot leg 29 & 30 Reflood assist bypass piping 31 Pressurizer surge line 32 Pressurizer 33 Piping connecting the steam generator secondary steam dome to the steam relief valve 34 ECC injection line 35 Steam generator secondary.downcomer 36,38,39,40 & 41 Steam generator secondsry shroud region 37 Containme nt 42 Steam ge:.erator time dependent volume I

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I Table 1 (cont.) RELAP4 Blowdown System Model Description I

Ilea t Slab Description 1,2 & 3 Nuclear Lore 4

Steam generator inlet plenum wall 5-12 Steam generator tubes 13 Lower core support structure 14 & 15 Upper core support structure 16 Upper head I

17 Steam generator outlet plenum wall 18 & 19 Core barrel and flow skirt 20 Lower plenum vessel' wall i

21 & 22 Intact loop hot leg piping 23,24 & 25 Intact loop cold leg suction piping 26 & 27 Intact loop cold leg discharge piping

[

28,29,30 & 31 Broken loop hot leg piping 32 & 33 Broken loop cold leg piping 34 & 35 Reflood assist bypass piping u

36 & 37 Steam generator secondary vescel wall i

28 & 39 Reactor vessel filler I

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Table 2 L3-6 Ppst > Test Analysis Predicted Pris1ry System Coolant Mass Balances at 2000 seconds I

lbm Initial Inventory 12,370 Total HPSI 2,245 Total PCPI 432.2 I

Total Break 13,700 Final Inventory 1347.2 I

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ML20030B499

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4.0 CONCLUSION

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