NUREG/IA-0040, 'Boil-Off Experiments with the EIR-NEPTUN … · 2012-11-21 · A simplified flow diagram of the NEPTUN facility is shown in Figure 1. NEPTUN consists of three main

NUREG/IA-0040EIR-Bericht Nr. 629

InternationalAgreement Report

Boil-Off Experiments with theEIR-NEPTUN Facility:Analysis and Code AssessmentOverview ReportPrepared byS. N. Aksan, F. Stierli, G. Th. Analytis

Laboratory for Thermal-HydraulicsPaul Scherrer Institute (PSI)5232-Villigen PSISwitzerland

Office of Nuclear Regulatory ResearchU.S. Nuclear Regulatory CommissionWashington, DC 20555

March 1992

Prepared as part of-The Agreement on Research Participation and Technical Exchangeunder the International Thermal-Hydraulic Code Assessmentand Application Program (ICAP)

Published byU.S. Nuclear Regulatory Commission

NOTICE

This report was prepared under an international cooperativeagreement for the exchange of technical information.: Neitherthe United States Government nor any agency thereof, or any oftheir employees, makes any warranty, expressed or implied, orassumes any legal liability or responsibility for any third party'suse, or the results of such use, of any information, apparatus pro-duct or process disclosed in this report, or represents that its useby such third party Would not infringe privately owned rights.

Available from

Superintendent of DocumentsU.S. Government Printing Office

P.O. Box 37082Washington, D.C. 20013-7082

and

National Technical Information ServiceSpringfield, VA 22161

NUREG/IA-0040EIR-Bericht Nr. 629

InternationalAgreement Report

Boil-Off Experiments with theEIR-NEPTUN Facility:Analysis and Code AssessmentOverview ReportPrepared byS. N. Aksan, F. Stierli, G. Th. Analytis

Laboratory for Thermal-HydraulicsPaul Scherrer Institute (PSI)5232-Villigen PSISwitzerland


March 1992

Prepared as part ofThe Agreement on Research Participation and Technical Exchangeunder the International Thermal-Hydraulic Code Assessmentand Application Program (ICAP)

Published byU.S. Nuclear Regulatory Commission

NOTICE

This report is based on work performed under the sponsorship of the

Swiss Federal office of Energy. The information in this report has

been provided to the USNRC under the terms of the International

Code Assessment and Application Program (ICAP) between the United

States and Switzerland (Research Participation and Technical

Exchange between the United States Nuclear Regulatory Commission

and the Swiss Federal Office of Energy in the field of reactor

safety research and development, May 1985). Switzerland has

consented to the publication of this report as a USNRC document in

order to allow the widest possible circulation among the reactor

safety community. Neither the United States Government nor

Switzerland or any agency thereof, or any of their employees, makes

any warranty, expressed or implied, or assumes any legal liability

of responsibility for any third party's use, or the results of such

use, or any information, apparatus, product or process disclosed

in this report, or represents that its use by such third party

would not infringe privately owned rights.

Contents

Acknowledgements .......................................... 1

Summary 3

1 Introduction 4

2 Description of the NEPTUN Test Facility 52.1 General Description . ............................... 52.2 Test Section ....... ................................... 5

2.2.1 The Test Bundle ................................... 52.2.2 Octagonal Housing and Pressure Vessel ..................... 62.2.3 Steam Flow and Carry-over Measurement System .............. 72.2.4 Regulation of the Test Pressure ......................... 7

2.3 Water Loop ......... ................................... 72.4 Auxiliary Systems ........ ................................ 8

2.4.1 Fresh steam Supply System ............................. 82.4.2 Data Acquisition System ............................. 82.4.3 Nitrogen Supply System ....... ........................ 8

3 Experimental Procedure 9

4 Experimental Results and System Response 114.1 System Response-Base Case, Experiment 5007 ..................... 114.2 System Response-Effect of Core Power .......................... 124.3 System Response-Effect of System Pressure ...................... 124.4 System Response-Experiment Repeatability ...................... 12

5 Evaluation of the External Surface Thermo-couple Response 12

6 Code Assessment 146.1 Assessment of the frozen version of TRAC-BD1/MOD1 and problem areas 146.2 Model Improvements ........ .............................. 15

7 Conclusions 17

References 19

Tables 21

Figures 23

iii

1

Acknowledgements

The authors are grateful to the personnel of the experimental team (F. Stierli,H. Griitter, E. Frei and L. Voser) at EIR who provided the NEPTUN facility andalso the experimental data.

During the course of this work, encouragement and support given by ProfessorG Yadigarogli and G. Varadi are gratefully acknowledged.

This report is prepared as an account of work partly sponsored by the SwissFederal Nuclear Safety Authority (HSK) of the Swiss Federal Office of Energy(FOE), within the activities of the International Code Assessment Program (ICAP)of US Nuclear Regulatory Commission (NRC).

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Summary

The NEPTUN data discussed in this report are from core uncovery (boil-off) experi-ments designed to investigate the mixture level decrease and the heat up of the fuel rodsimulators above the mixture level for conditions simulating core boil-off for a nuclearreactor under small break loss-of-coolant accident conditions.

The first series of experiments performed in the NEPTUN test facility consisted often boil-off (uncovery) and one adiabatic heat-up tests. In these tests three parameterswere varied: rod power, system pressure and initial coolant subcooling. The repeata-bility of the experiments was also demonstrated.

The NEPTUN experiments showed that the external surface thermocouples do notcause a significant cooling influence in the rods to which they are attached under boil-offconditions. The reflooding tests performed later on indicated that the external surfacethermocouples have some effect during reflooding for NEPTUN electrically heated rodbundle. Peak cladding temperatures are reduced by about 30-40°C and quench timesoccur 20-70 seconds earlier than rods with embedded thermocouples[1]. Additionally,the external surface-thermocouples give readings up to 20 K lower than those obtainedwith internal surface thermocouples (in the absence of external thermocouples) in thepeak cladding temperature zone.

Some of the boil-off data obtained from the NEPTUN test facility are used forthe assessment of the thermal-hydraulic transient computer codes. These calculationswere performed extensively using the frozen version of TRAC-BD1/MOD1 (version

22). A limited number of assessment calculations were also done with RELAP5/MOD2(version 36.02). In this report the main results and conclusions of these calculations arepresented with the identification of problem areas in relation to the models relevant toboil-off phenomena. On the basis of further analysis and calculations done, changingsome of the models such as the bubbly/slug flow interfacial friction correlation whicheliminate some of the problems are recommended.

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1 Introduction

The drying out of a nuclear reactor core by boiling off the coolant inventory is known as acore uncovery event. The accident that happened at the Three Mile Island Unit-2 (TMI-2) Plant on March 29, 1979 can be classified as a small break loss-of-coolant accident

(SBLOCA) followed by core uncovery. The flow of the core was blocked and the coolantevaporated through the stuck-open pressure relief valve causing gradual depletion ofthe core coolant inventory. Part of the core was uncovered and extensive damage tothe fuel rods was done. After the occurence of this accident,' increased attention hasbeen paid in Light Water Reactor (LWR) Safety analysis to low flow and intermediate

or low pressure transients which, if no remedial measures are taken, may sequentiallylead to uncovery, overheating and damage of the core. When the reactor power is atdecay heat levels and the coolant entering the core is subcooled, uncovery of the coreis likely to occur only at relatively low coolant mass flow rates. Thus, hydraulic andheat transfer mechanisms associated with cooling of fuel rods in a pool of water withoutexternal circulation need to be better understood. In particular, there is an obvious needfor assessing thermal hydraulic safety codes as far as their ability to correctly predictthe level swell (or expansion of the boiling pool) and the fuel rod temperatures in theuncovered region. Consequently, considerable effort is being spent not only in improvingand-extending the available transient thermal hydraulic codes, but also in performingcarefully controled experiments in test facilities; the results of these simulations can bedirectly utilized for assessing the predicting capabilities of these codes or even developingnew models, hence giving a direct feed-backto the code-developers.

At the Swiss Federal Institute for Reactor Research (EIR), the NEPTUN facility[2] and rod bundle were originally designed for low pressure (_5 5 bars) reflood exper-iments whose aim was to study the heat transfer characteristics between the rods andthe coolant. Additionally, a number of core-uncovery (boil-off) experiments have beenperformed to investigate the mixture level decrease and resulting fuel rod heat-up abovethat level that may occur in a PWR during small and intermediate break LOCAs. Theseboil-off tests have been performed using a variety of initial parameters (eg. rod power,system pressure, coolant subcooling etc.).

In these experiments it was also possible to evaluate the accuracy and cooling influ-ence of the external surface thermocouples (similar to those used in the LOFT facility atthe Idaho National Engineering Laboratory) over a range of power and system pressures.

This report summarizes the NEPTUN boil-off experiments as well as comparisonsbetween experimental data obtained from the tests with corresponding predictions ob-tained using the best-estimate thermal hydraulic code TRAC-BD1[8] .

The NEPTUN system configuration and boil-off tests and related test matrix arediscussed in Sections 2 and 3, respectively. Experimental results and system response

are presented in Section 4. Section 5 summarizes the results of NEPTUN boil-off exper-

iments with regard to external thermocouple response. Section 6 discusses the resultsof code assesssment calculations. Conclusions related to NEPTUN boil-off experiments

5

are presented in Section 7.

2 Description of the NEPTUN Test Facility

2.1 General Description

A simplified flow diagram of the NEPTUN facility is shown in Figure 1.NEPTUN consists of three main parts:

* The test section, including

- a measuring system for the liquid carry-over rate and for the steam expelledduring the experiment,and

- a backpressure control system.

* A water loop to bring the water to the test conditions.

* Several desired inlet or initial auxiliary systems to maintain normal operatingconditions.

2.2 Test Section

2.2.1 The Test Bundle

The test bundle consists of 33 electrically heated rods and 4 unheated guide-tubes placedin an octagonal housing as shown in Figure. 2. . The NEPTUN bundle arrangementcorresponds to a section of the LOFT nuclear fuel bundle.

The length of the bundle is subdivided into 8 equally spaced measurement levels(Fig. 3). At these levels local fluid temperatures, pressures and pressure differenceswithin the test section are measured.

A detailed description of the components in the test bundle follows:

The heater rods have a heated length of 1680 mm. Figure 4 shows the axial choppedcosine-type power distribution used. The axial power distribution for the LOFT nuclearfuel and the SEMI-SCALE heater elements are also given in this figure for comparison.Details of the NEPTUN electrical heater rods and their main specifications are shownin Figure 5.

. The cosine power distribution is obtained by a kanthal heater rod of axially changingdiameter in the center of the heater element. The electric current, flowing from thebottom to the top of the element, is led back to the bottom by a noncentric coppertube. The kanthal rod is electrically insulated by a boron nitride layer; the copper tubewhich is placed between two thin inconel tubes is also insulated by an A1203 plasma

6

coating. The canning consists of 2 thin-walled inconel tubes. The heater rods areequipped with 4 or 8 thermocouples which are situated in small notches between thetwo outer inconel tubes. Several drawing processes during the assembly of the heaterelement assure good heat transfer from one layer to the next.Axial heat conduction calculations did show that the axial heat profile is not seriouslydisturbed by the good heat conductivity of the copper tube [4].

Guide tubesIn the NEPTUN bundle there are 4 unheated guide-tubes. Their positions are shownin Figure 2. The outer diameter of the guide-tubes is 13.87 mm and is slightly largerthan that of the heater rods.All of the 4 guide-tubes are equipped with 8 thermocouples to measure the temperatureof the guide-tube wall.

Spacer gridsThe spacer grids are of the same design as in the LOFT nuclear core. There are 5 spacergrids, equally spaced, within the NEPTUN test section. Their axial positions are shown

in Figure 3.

2.2.2 Octagonal Housing and Pressure Vessel

In order to obtain a low heat capacity, the housing has been designed with a wall thick-ness of only 2 mm. Because the thin wall cannot stand too high pressure differencesat high temperatures, the housing is surrounded by a pressure vessel containing pres-surized nitrogen. The space between the housing and the pressure vessel is also filledwith an insulating (low heat capacity) material. The pressure in this place is controlledby a pressure regulating system and is held automatically at approximately the samepressure as inside the housing.

The shape of the housing (Figure 2) is chosen in such a way that the parasitic

flow area does not exceed 9 percent of the actual flow area in a square pitched 37 rodbundle. The housing is fabricated from inconel 600 tubing which is formed into therequired shape by several drawing processes.

Two tubes, displaced by 180 degrees, are welded radially onto the housing at each

measurement level (Figure 2). Each one of these tubes is connected with the test spaceby a boring of 2 mm diameter through the wall of the housing. The tubes can be usedto measure pressures, pressure differences or fluid-temperatures.

The axial and azimuthal temperature distributions of the outer surface of the housingand axial and radial temperature distributions in the insulating material between thehousing and the pressure vessel are measured by up to 31 thermocouples. Details are

shown in Figure 6.

ý't .'

7

2.2.3 Steam Flow and Carry-over Measurement System

The two-phase mixture which is expelled from the test section is separated into steamand water. The steam flow is measured with a turbine flow meter in the steam exhaustline, while the entrained water is collected in the carry-over tank.

Steam-water separator and carry-over tank (Figure 7)Directly above the upper end of the test section, there is a diverting plate which deflectsthe steam-water mixture expelled from the test section into horizontal direction. Dueto the large cross sectional area of the steam-separator, the steam moves slowly towardsthe steam exhaust line. The water falls down to the bottom of the separator and flowsthen into the carry-over tank in which the carry-over rate is measured.

Exhaust steam line(Figure 1)After leaving the separator, the steam flows through a flow measuring system consistingof:

- an electrically heated tube to dry and superheat the steam for avoiding any con-densation (item 12),

- a turbulence promoting plate for equalizing the temperature profile in the flow(not shown),

- a flow straightener to suppress any rotational component of the flow (not shown),

- a tap for pressure measurement,

- a turbine flow meter (item 8),

- sensors to determine the steam temperature.

2.2.4 Regulation of the Test Pressure

The experiments should be run at a test section pressure as constant as possible, in spiteof time dependent steam production during the boil-off phase. Therefore a regulatingsystem is necessary. It consists of a V-ball valve (Figure 1, item 10) and a regulatingunit, which actuates the valve in such a way, that the desired pressure level is keptconstant. The actual pressure signal is supplied to the regulating unit from the steamwater separator chamber.

2.3 Water Loop

The water is brought up to the required inlet or initial conditions (temperature, pressure,flow rate) in a closed water loop (fig 1). The following components are available to obtainthe desired water conditions:

8

* a water supply tank with an electric heater (item 4),

" a cooler (item 5),

* a pump (item 6),

" a regulating valve (item 13),

" a motor valve (item 14).

To avoid oxidation of the heater rods as much as possible and to avoid any depositionon hot surfaces, demineralized water with very low oxygen content is used. This lastcondition is obtained by boiling and degassing the water in the water supply tank atatmospheric pressure.

2.4 Auxiliary Systems

2.4.1 Fresh.steam Supply System

A steam boiler (fig.1, item 7) produces-steam by boiling demineralized, oxygen freewater. Prior to heating-up the heater rods, steam is fed into the steam-water separator:

" to purge the whole test circuit,

" to build up and maintain the desired test pressure.

2.4.2 Data Acquisition System

The data acquisition system is the primary data collecting system and consists of aHP-2100 computer and associated equipment. The system can record 300 channels ofanalog input data representing bundle and system temperatures, bundle power, flows,and absolute and differential pressures. Each data channel is recorded at least onceevery 2 seconds. The digitized data are stored on magnetic tape. The data reductionand processing is carried out at the computer center of EIR.

2.4.3 Nitrogen Supply System

Nitrogen is used:

- to adjust the pressure in the pressure vessel according to the pressure in the testsection, in order to avoid an overloading of the housing,

- to maintain a slight overpressure in the main test loop during shut down. This isa preventive action for keeping the oxygen content in the test section as low aspossible.

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3 Experimental Procedure

This chapter is a summary of significant instructions which were followed during exper-iments 5000-5011.

In each experiment the following 3 parameters defined the test matrix:

" the test section pressure pand the corresponding saturation temperature T,•t,p

" the initial temperature of the water in the T1test section and corresponding saturation pressure psat,t

" the heater-rod power.

In order to establish the desired initial test conditions, a series of operations wereconducted prior to the start of the experiment. The main operating instructions aredescribed below:

1. 1-2 Hours prior to the start of a boil-off experiment:Degas the demineralized water in the water supply tank by boiling at atmosphericpressure.Circulate the water in the water loop and adjust the parameters to the desiredvalues:

(a) The pressure in the water supply tank to a value approximately 0.5 bar higherthan pat,1 by boiling water at a small power rate and bleeding an appropriateamount of vapour.

(b) the temperature near the test section inlet valve to Tt by adjusting the flow-rate (and therefore the heat losses in the piping system) to an appropriatevalue.

(c) Heat up the fresh steam generator to a temperature corresponding to a pres-sure of approximately 8 bars.

(d) Heat up the main components in the contact with steam by means of electricstrip heaters:

the steam-water separator,

the entrainment tank, to T,,t,p

the exhaust steamline,the superheater to 250-300 0C

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2. Purge all pressure transmission lines of the differential pressure measuring systemwith cold, degassed water. Vent the differential pressure cells.

3. a) Determine the calibration coefficients of the test section differential pressuretransducers by means of a two point method a assuming a linear characteristic,

b) Set the zero of the absolute pressure transducers at atmospheric pressure.

All other pressure transducers are calibrated prior to experiment 5000 only.

4. Set the pressure control system to maintain a constant pressure psat,t in the testsection.

5. Increase the test section temperature (housing, heater rods, guide tubes) to theinitial test wall temperature T, by feeding steam from the steam generator intothe steam-water separator until the pressure in the test-section reaches Psat,t-

After reaching psat,t, purge the test section during a few minutes with a smallsteam flow.

As soon as temperatures of the heater rod and housing have stabilized at T,:fill the test section with water of temperature Tt from the water loop. The testsection pressure is held constant by the pressure control system.

6. Set the pressure control system to maintain a constant pressure p in the testsection.

7. Increase the pressure in the test section to p by feeding again steam into thesteam-water separator.

Maintain a continuous steam flow from the steam generator through the steam-water separator and the exhaust steam line.

After reaching p purge the entrainment tank several times with steam to stabilizethe wall temperature at T,,t,p.

8. As soon as the wall temperatures have stabilized at Tar,p, increase the steam flowto a value at which both the fresh steam turbine and the exhaust steam turbineare operating.

9. Switch the data acqtuisition computer into the fast-scanning mode.

10. When all of the specified initial conditions are established, start the experimentby applying the desired electric power to the heater rods.

acompletely flooded test section and empty test section

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11. Keep the rod temperatures under observation. (10 digital displays were availablefor monitoring rod temperatures of special interest). As soon as the maximumrod temperatures reach -850C shut off the electric power to the heater rods.

12. Shut down NEPTUN by:- flooding the test-section with water from the water loop,- draining the test section and the entrainment tank,- switching off all heaters, pumps, etc.

During the cooldown of NEPTUN nitrogen is automatically fed into the main testloop to maintain a pressure of approximately 1.1 bar.

4 Experimental Results and System Response

The ten core boil-off experiments performed are summarized in Table 1. Three parame-ters were varied-rod power, system pressure, and initial coolant subcooling. Rod powerlevels were chosen to represent the nuclear decay power. System pressure was variedover the 1-5 bar range possible in the NEPTUN facility.

Experiment number 5007 (see Table 1) was chosen as the "base case" because it wasconducted at higher pressure and intermediate power. However, several experiments willbe discussed to show the difference in the NEPTUN system response due to varying rodpower, system pressure, initial water subcooling, and finally, to demonstrate experimentrepeatability.

4.1 System Response-Base Case, Experiment 5007

Figure 8 presents an overlayb of the core power history, core fluid level as measured bythe core total Ap measurement, and typical responses of the heater rod thermocouplesfor the base case. Notice that the power is increased at about 50 s, and that a rapiddrop in the core fluid level occurs at about 100 s. This delay time between initial powerand the rapid initial liquid level decrease is a result of heating the. subcooled water tothe saturation temperature. Shortly after the water reaches the saturation temperature(between 100 and 110 s), large voids are formed due to vapor generation, expelling someof the liquid from the core. After this initial liquid swell, the core liquid continues tobe slowly boiled off as shown by the decreasing core Ap in Figure 8. The claddingthermocouples dry out and heat up with differing heat-up rates for each axial elevation.At -,800 s the power is shut off and the rods are allowed to cool down before the systemis reflooded.

bThe system response overlays are presented to compare selected general system response data

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4.2 System Response-Effect of Core Power

The effects of increasing the core power (experiment 5006) and decreasing core power(experiment 5008) on the boil-off response are presented in Figures 9 and 10, respec-tively, which can be compared directly to the base case (5007) response shown in Fig-ure 8. The system responded as expected in each case; increased core power resulted inmore rapid boil-off and cladding heatup, and decreased power resulted in the oppositetrends.

4.3 System Response-Effect of System Pressure

The effect of lowering the system pressure from 5 to 1 bar (experiment 5002) on thecore liquid level and cladding temperatures is shown in Figure 11. The lower systempressures result in more rapidly decreasing liquid levels and rod dryouts ranging from100-200 s earlier than experienced for the base case (Figure 8).

4.4 System Response-Experiment Repeatability

The experiment was very repeatable for both high- and low-pressure conditions. Ex-periments 5001 and 5002 were repeat experiments at low pressures. Figure 12 showsthe system response for the repeat conditions (experiment 5001) for comparison toFigure 11, discussed in the previous section. Experiments 5006 and 5009 are repeatexperiments at high pressure and comparison is shown in Figures 9 and 13.

5 Evaluation of the External Surface Thermo-couple Response

One of the objectives of the boil-off experiments in the NEPTUN facility was to obtainexperimental data for assessing any perturbing effects of the external surface, thermo-couples used in LOFT, during simulated small-break core uncovery conditions. Prior tothe NEPTUN Experiments, it was hypohtesized that the external surface thermocou-ples might cause additional selective cooling of the rods, which would result in delayeddryout for a slow core uncovery experiment in LOFT. Also, the added increase in surfacearea for heat transfer (fin effect) might result in additional atypicalities.

The first series of experiments performed in NEPTUN consisted of eleven tests,ten boil-off and one adiabatic heat-up test. In these tests, three parameters were var-ied: rod power, system pressure and initial coolant subcooling as already describedin chapter 4. The effects of the external surface thermocouples were detemined bycomparing the cladding temperatures (as measured by the external LOFT-type surfacethermocouples) to cladding temperatures from thermocouples within the cladding ofthe NEPTUN heater rods[l]. Figure 2 shows a schematic of both the external surfacethermocouples (LOFT-type) and NEPTUN embedded thermocouple configuration used

13

for the experiments discussed in this report. Note that there is only one active externalsurface thermocouple on each of the five LOFT-type rods, the other external surfacethermocouples are replaced by dummy elements.

Overlay plots of the thermocouple responses for many different thermocouples ateach axial elevation are contained in the appendices of reference 5 for each experiment.These plots indicate that the readings of the external surface thermocouple is well withinthe response spread of the internal thermocouples. In this report, as bounding cases,experiments 5007 (base case) and 5011 corresponding to small and intermediate breakLOCA decay heat levels, respectively, are discussed.

Overlay plots of the thermocouple responses at level 4 corresponding to the maxi-mum linear heat generation position on the heaters are shown in figures 14 and 15 forexperiment numbers 5007 and 5011 respectively. Notice that a systematic lower temper-ature is measured by the external surface thermocouples. This temperature differencecan be taken as an estimate of the cooling effect of the external surface thermocouplesand is less than 20 K. Notice also in figures 14 and 15 that there is less than 5-10 sdifference in the initial dry-out times for all level 4 thermocouples, both embedded andexternal.

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6 Code Assessment

The NEPTUN boil-off experimental data were used for assessing the thermal-hydraulictransient computer codes TRAC-BD1/MOD1 and RELAP5/MOD2. The assessmentcalculations performed with TRAC-BD1/MOD1 will be given in more detail in thenext subsections. During the early phase of the assessment of the RELAP5/MOD2code, some simulation and calculational difficulties were encountered for boil-off casese.g. very large discrepancy in calculating the amount of expelled water out of thetest section as shown in figure 16 [11]. Further calculations were not performed withRELAP5/MOD2, until the reasons for such discrepancies were identified. Two otherattempts using RELAP4/MOD6 and RELAP5/MOD1 were also not very successful [13].The RELAP4/MOD6 calculations could be performed until the onset of nucleate boilingand as soon as vapor was produced, very large pressure spikes were observed, resulting intime consuming and costly calculations. RELAP5/MOD1 which employes five-equationhybrid model calculated about 100 seconds earlier critical heat flux occurance withrespect to RELAP5/MOD2 and the amount of expelled water out of test section waseven more over-predicted.

6.1 Assessment of the frozen version of TRAC-BD1/MOD1and problem areas

A number of NEPTUN boil-off experiments have been utilized for assessing the predict-ing capabilities of the thermal-hydraulics transient analysis code TRAC-BD1. Origi-nally, the experiments were analyzed with version 12 of TRAC-BD1 [31 and the problemareas were identified [6,7]. Subsequently, five of these experiments were re-analyzed byusing a frozen version of the code (version 22) commonly known as MODI [8].

Since the models related to the dominant physical phenomena in these core uncoveryexperiments are the same in both versions of the code and the problem areas identifiedwith MOD1 were almost the same with the ones of version 12, we shall concentrate onthe results obtained by employing TRAC-BD1/MOD1. For more details, the interestedreader is referred to a series of reports [6,7]; here we shall outline our findings related tothe problem areas as well as the model improvements already implemented in MODI.One of these improvements is already included in the BF1 version of the code recentlyreleased.

Four boil-off experiments at 5 bar and one at 1 bar were analysed using the frozenversion of TRAC-BD1/MOD1 and experiments are separately summarized in Table 2. Anumber of numerical problems have been revealed in the course of the analysis of theseexperiments with TRAC and have been extensively analysed and reported elsewhere[6,7]; here, we shall restrict our attention on the problem areas of the code related tothe actual physical modeling of the phenomena taking place.

Comparison of measured and calculated collapsed liquid level (CLL) and peak axialpower level rod surface temperature histories for the five experiments are shown in

15

Figures 18 to 22. One can readily draw the following conclusions regarding the predictingcapabilities of the code:

(a) TRAC-BD1/MOD1 underpredicts the CLL histories and hence, predicts an earlierCHF than the measurements show. Clearly, the code overpredicts the amount ofwater expelled from the test section. These differences are more pronounced forthe 1 bar experiment.

(b) Generally, TRAC-BD1/MOD1 predicts an earlier CHF than the measurementsshown; hence, the sudden expulsion of water from the test section is predicted tooccur earlier [6,7J. Also, the predicted rod surface temperatures during nucleateboiling are 8-15 K below the measured ones.

(c) It was noticed that after the rod power was turned off, the slopes of the predictedand measured rod surface temperatures were different, indicating that the calcu-lated heat transfer coefficient in this region was overpredicted. This can be seenin Fig. 20 for Exp. 5007. This was changed as we shall discuss in due course; allthe boil-off runs to be reported in this report were made with the code versionincorporating this change.

6.2 Model Improvements

As Figs. 18-22 show, the main problem of TRAC-BD1 is that it overpredicts the amountof water expelled in the boil-off tests. The origin of this was traced back to the ratherhigh interfacial shear calculated by the interfacial friction correlation used for bub-bly/slug flow. This correlation although appropriate for tubes, has recently been shownnot to be suitable for rod bundles [9]. The interfacial shear force per unit volume fiin TRAC-BD1 for the bubbly/slug flow regime is based on the following vapor driftvelocity correlation

Vd = V2 I{ (PI2P.)}1/ 4 (1)

Starting from this expression, it can readily be shown [8,10] that

f, = pja(1 - a)' I5 _ C0 3-fi 9 - C 13 I , vCl - Col') (2)

where a is the surface tension, g the gravity constant,

C, 1 - aC0(301=- 1 -a (3)

and Co ý- 1.3. Based on the work of Bestion [9], we implemented in TRAC-BDI a newbubbly/slug f, correlation suitable for rod bundles; it is based on the following vapordrift velocity correlation

16

Vd = 0 . 12 4{ g(P -- P)DH}/ 2 (4)

where DH is the channel hydraulic diameter; this results in the following expression forf/"

fi = 65a(1-a HC) 3 pg I ClVg - CoVI I (Cvg - CoVy) (5)

Figures 23-27 show the comparison of the measured and calculated CLL's (with thenew TRAC-BD1/MOD1 version) and peak axial power level rod surface temperaturehistories for the five boil-off experiments. Except in the 1 bar case, for which thecode still underpredicts the CLL history, there is now excellent agreement betweenmeasurements and predictions. The code developers have already implemented thisnew fi correlation in the new code version TRAC-BF1.

We could not trace the origin of the earlier transition to nucleate boiling predictedby the code; though, recent work [11] has shown that since the thermocouples are alittle below the rod surface, an oxide layer having a thickness of 25 prm on the surfacecould in fact increase the thermocouple readings by as much as 12 K. This is a plausibleexplanation of the differences between measured and predicted rod surface temperaturesduring nucleate boiling.

Finally, the problem of overprediction of heat transfer after the power was turnedoff was traced back to the steam cooling logic of the code and since this problem wasnot encountered when analyzing the boil-off experiments with version 12 of the code[6,7], the steam cooling heat transfer logic of version 12 was re-introduced in MOD1.

Specifically, for steam cooling, in MODI the following heat transfer coefficients aredefined [8]

hv•iam = 4 (6)m DHf

= 0.023(Ret) 0 8(Pr)0-33K,,/DIr (7)

h,,, = 0.13( Gr.Pr)°'333 / D•i (8)

where all the symbols have their usual meaning and

Pg Il T. - q D1,3Grv.= 1-g- T DH (9)

In MOD1, the following selection logic exists for h, if a > 0.999:

h= = hutur(10 (10)

17

If h. < hv,tam, h. = hv,,,am (11)

If hi,<h,,,< hv i = hv,,c (12)

This was modified as follows (for a > 0.999):

h,, = MAX{hv,,c, hi,tr } (13)

Comparison between measured and predicted peak axial power rod surface temperaturehistories for Exp. 5006 are shown in Fig. 28a (standard MOD1) and 28b (modified asabove).

Based on the experience gained from TRAC-BD1 modifications, the bubbly and slugflow regime interfacial friction correlation used in CATHARE code for bundle geometrieswas implemented into RELAP5/MOD2. As it can be seen from figures 16 and 17, theresults of the calculated entrained water and cladding surface temperature are very wellcomparable with the experimental data of experiment 5007.

7 Conclusions

The NEPTUN experiments have provided thermal-hydraulic data simulating nuclearreactor core boil-off conditions at low pressure (1-5 bar). The data obtained from thesetests proved to be useful in assessing the modeling capability of available computercodes.Analysis of the experimental boil-off data indicate that:

" increasing core power resulted in more rapid boil-off and cladding heat-up, whiledecreasing power resulted in the opposite trends, as expected

" the lower system pressures resulted in more rapid decrease of liquid levels andfaster rod dry-outs relative to the base case

* dry-out times of the internal and external surface thermacouples were within10 seconds of each other at any axial elevation for all rods in the bundle. Thecladding external surface thermocouples measure the cladding temperatures thatwould have been measured in their absence within 0 to -20 K.

Analysis of a number of NEPTUN boil-off experiments and comparisons with TRAC-BD1/MOD1 predictions showed that:

e the collapsed liquid level history is underpredicted and consequently, CHF occursearlier than in the experiments. Clearly, TRAC overpredicts the amount of waterexpelled from the test section.

18

" Generally, earlier incipience of nucleate boiling is predicted; and recent inves-tigations indicate that differences between measured and predicted rod surfacetemperatures during nucleate boiling can be due to the formation of an oxidelayer around the electrical heater rods.

" After the rod power was turned off, the slopes of the predicted and measured rodsurface temperatures were different, indicating that the calculated steam coolingheat transfer coefficient was overpredicted.

To improve the prediction capability of TRAC-BD1/MOD1 the following main modifi-cations were introduced:

" An alternative bubby/slug interfacial shear correlation, more appropriate for bun-dles and used in the CATHARE code, is implemented in the code. As a result ofthis change, the collapsed liquid level histories are correctly predicted by decreas-ing the interfacial friction in this flow regime.

" The steam cooling heat transfer logic used in version 12 is re-introduced in MOD1,specifically to eliminate the differences during the steam cooling phase after thepower was turned off.

19

References

[1] S.N. Aksan, et al. "NEPTUN Bundle Boil-Off and Reflooding Experimental Pro-gram Results". Presentation at the Tenth Water Reactor Safety Research Informa-tion Meeting, Gaithersburg, Maryland, USA, NUREG/CP-0041, Vol. 1, (1982)

[2] H. Grfitter, F. Stierli, S.N. Aksan, G. Varadi, "NEPTUN Bundle Reflooding Ex-periments: Test Facility Description" EIR-Report No. 386, (1980)

[3] J.W. Spore et al.," An Advanced best Estimate Computer Code Program for Boil-ing Water Reactor Loss-of-Coolant Accident Analysis, Vol. 1, 2, 3. NUREG/CR-2178 (1984)

[4] S.N. Aksan, S. Gfintay, G. Varadi, "Analytical Investigations of the Thermal Be-haviour for NEPTUN Heater Elements (with Cosine Heat Generation) used inReflooding Experiments"; Proceedings of the International Symposium on FuelRod Simulations - Developments and Application, Gatlinburg, Tennessee, CONF-9-1091, (1980)

[5] E.L. Tolman and S.N. Aksan: "Summary Results of the NEPTUN Boil-Off Ex-periments to investigate the Accuracy and Cooling Influence of LOFT CladdingSurface Thermocouples", EGG-LOFT-5554, EG&G International Report, (1981)

[6] G.Th. Analytis and S.N. Aksan, "TRAC-BD1 Assessment under Severe AccidentBoil-Off Conditions", Fifth International Meeting on Thermal Nuclear ReactorSafety, Karlsruhe, 9/9-13/9, Proceedings Vol. 3, p. 1821, (1984)

[7] G.Th. Analytis and S.N. Aksan, "Trans. Amer. Nucl. Soc., 47, 493, (1984)

[8] D.D. Taylor et al., "TRAC-BD1/MOD1: An advanced best Estimate ComputerProgram for Boiling Water Reactor Transient Analysis, Vol. 1, 2, NUREG/CR-3633, (1984)

[9] D. Bestion, "Interfacial Friction Determination for the ID-6 Equations Two-FluidModel used in the CATHARE Code", presented at the European Two-Phase FlowGroup Meeting, Southampton, England (1985)

[10] G. Th. Analytis, Trans. Amer. Nucl. Soc., 52, 481, (1986)

[11] G.Th. Analytis, M. Richner, M. Andreani, S.N. Aksan, "Assessment and Uncer-tainty Identification for RELAP5/MOD2 and TRAC-BD1/MOD1 Codes underCore Uncovery and Reflooding Conditions", 1 4th Water Reactor Safety Informa-tion Meeting, Gaithersburg, Maryland, USA, (1986)

20

[12] G. Th. Analytis, S.N. Aksan, F. Stierli, G. Yadigaroglu, "Dynamics of Core VoidingDuring Boil-Off Experiments", 4 th Miami International Symposium on Multi-PhaseTransport and Particulate Phenomena, Miami Beach, Florida, USA (December1986)

[13] M. Andreani, "Brief Notes on the Simulation of the NEPTUN Boil-Off Experiment5007 by means of the RELAP4/MOD6 and RELAP5/MOD1 (cycle 25) ComputerCodes", Letter to S.N. Aksan, 24.11.1986

21

Table 1: Summary of NEPTUN Boil-Off Experiments

Rod InitialBundle Peak System Coolant Initial

Experiment Power Power Pressure Temperature SubcoolingNumber (kW) (kW/m) (bar) ("C) K Comments

5000 24.6 0.744-4.6 1 100 0 Power was too highfor the first test; thedata were not evaluatedfrom this experiment

5001 24.6 0.744 1 100 0 Repeat experiment at lowpressure

5002 24.6 0.744 1 100 0 Repeat experiment at lowpressure

5004 24.6 0.744 5 120 32 Effects of changing rodpower and initialcoolant subcooling

5005 42.1 1.276 5 120 32 Effects of changing rodpower and initialcoolant subcooling

5006 42.1 1.276 5 140 12 Effects of chaning rodpower at high systempressure

5007 24.6 0.744 5 140 12 Effects of changing rodpower at high systempressure

5008 10.5 0.319 5 140 12 Effects of changing rodpower at high systempressure

5009 42.1 1.276 5 140 Repeat of high-power,high-pressure experiment5006 with higher data-scanning rate

5011 75.1 2.276 5 112 38 High power test5012 42.1 1.276 5 steam steam Adiabatic heat-tip test

and, from 345 to 470 s

flooding with floodingvelocity of 15 cm/s andcoolant temperature of 74"C

22

EXPERIMENT NUMBER

5002 5006 5007 5008 5011

PRESSURE (BAR) 1 5 5 5 5

SUBCOOLING ( 0 K) 0 12 12 12 39

BUNDLE POWER (KW) 24.6 42.1 24.6 10.5 75.1

TABLE 2

23

9 I--

I10

J47...."

1 HEPliUN- TeSTSEcTION2 STE•M WATER SEPARATOR

3 CARRYOVER TANKWATERSUPPLY TANK

S WATER COOLER

6 rt7 STEM BOILER

8 TURBINE FL.0MICTER

9 QUICK ACTING VALVE

10 fRESSURE CONTROL VALVE

11 "ESSURE CONTROLLER

12 STEM SUPERHEATER

13 REGULATING VALVE

1.4 MOTOR VALVE

MIXAiC TEST Loop fWATER LOOP 4

I SMALL LINE

WATER QUALITY

14Y

?

z

Figure 1: Simplified NEPTUN Flow Diagram

24

SAPH~A

1 0

1

NEPTUNEXPERIMENT NR, 5000 1 5001

-740t Dufsdl Configuration..

-Measured Rod lemperatures

Healer Rods

4 4

43. 4A8A' n~dtyo.

WALD

1.2.3.4.7.6

Wets*t eucIten.I •

[L~444t 0-6"l~gt ow4

Ks Tcia mi5n w.. U'S .of

Gel-d1 Tut01

N,4 .3,fi MP-111n Uwcmos..).

vien ftom above

BENAU(no. ith

Figure 2: NEPTUN Bundle Cross-Section

25

NEPTUN

dimensions in (mm]

dimensions validao room temperalure

Figure 3: Axial Distribution of the Measurement Levels and the Spacers

26

measurement levels

I i I 2 3 I 1 5 1 618

1.50

1.01ý

- - -- LOFT

-.... SEMI SCALE

NEPTUN

L.

- - V

0

120[co

Icv~er end

Neptun,

0~ W e66 1/ -0.!5 - 1-08.e [Cos V66 X CE 143 .j-0,2412 . 1.339-cas izzi143 oX C6%O :t/ . Q241Z VL 15' 4

10 20 30 40 50 60 70 80 90 100 110 Izo 130 140 150 160 170 (co]

2

LOFT type spocers

3upper end 54

Figure 4: NEPTUN: Axial Power Distribution of the Heater Rods

0

0

0

0

0

0

0

HEATED LENGTH

0. DIAMETER

POWER DISTRIUTION

ROD POWER

AVERAGE HEAT FLUX

PEAK HEAT FLUX

PEAK LINEAR HEAT RATING

AXIAL PEAKING FACTOR

1680 MM

10.72 MM

CHOPPED COSINE

4.5 KW (MAX)

7.86 W/CM 2

12.4 W/CM 2

41.9 WICM

1.58

THERMOCOUPLES

Figure 5: NEPTUN: Electrical Heater Rods

28

PRESSURE VESSELO-REFERENCEAXIS

iI J HOUSING

180000 900 2700 360* B O: AZIMUTH

I 1TM- - 26 1 j ___11

MEASUREMENTLEVEL

62

69-- a a-

66 -167 68-a -- I N THERMO -

COUPLE

--I

a n._-70 72 73

74

180-

84-- - -

84- - -- 7-

86 87 88 90 91

94

6

5

FOR DETALSREFER TODRAWINGS:

0-171 057.8(HOUSING)

0-171 086 A(PRESSUREVESSEL)

2

1

Figure 6: NEPTUN (Experiment no. 5000-5007): Measured temperaturesat the outer surface of the housing and within the insulation

29

WIRE MESH SEPARATOR

DIVERTING PLATE

-TEST SECTION

HOUSING

PRESSURE VESSEL

Figure 7: NEPTUN: Steam Water Separator and Carry-Over Tank

800

700

600

• 500

01

:-

E400

CM

co 300

200

100ý

50

45

40

35

30

25 "0-0~

20"

15

10

200

180

160

140

C.01

100

80

--60

W0

-1405

0

5

1000

20

0500

Time (s)

Figure 8. NEPTUN system response--base case (test 5007--high systempressure, intermediate rod power).

800200

180

160

140

CL0

E0•

C3

03

L.

C)0C.

"c"120 E

c0100 =

0

80 _

tA)H

60

40

20

0500

Time (s)

Figure 9. NEPTUN system response--test 5006--high system pressure,high rod power.

800 50 200

700 ---- THeater rod power - 45 180TO level 5TO level 6TO level 3 40 160TO level 7

600 - TC level 8 35Tlevel 2and 4 140

.500 - - 305120 E

S25

~400 -oý-00 ° -100.z

CD 20-4OO-

-o 80~as 3001

15 0

rem- 10. - 605 -, 40o100-

0 - 200 -- ' 05 -20

0 100 200 300 400 500 600 700 800 900 1000Time (s)

Figure 10. NEPTUN system response--test 5008--hiqh system pressure, low rodpower.(Figure continued)

800 I I I! I I I I I I I I

700 1-

600

-am -- - -

- . -

Total test sectionAp(mixture level)

Heater rod powerTC level 5TC level 4TC level 6TC level 3TC level 7TC level 8TC level 2

'-a,

-a5

?..500

CD

'aC 300

-43

.•7 I • ._.•0 ...........

- ---- .. . .. - -.- -- 2 _.- - . - -

50

45

40

35

30

25-,0

15'0

20-0

15

10

5

0 -

-5 -00

160

200

180

140

1200.4C0

100

0

60

40

20

WA

200

100 F-

20

I0110C

I I I I I I I II I 00O 1100 1200 1300 1400 150 0 1600Time (s)

1700 1800 1900

Figure 10. (continued).

800

700

600

. 500

0.

2 4000)

'D"1012300

200

0

50 -

45 _

40

35

30

25,0

20-

0.

1-515

200

180

-160

140

120.9

.0180

0.

40

20

W

10

-5

0

-51000

-j00 100 200 300 400 500 600 700 800 900

Time (s)

Figure 11. NEPTUN system response--test 5002--low system pressure,intermediate rod power.

35

0

C

CCECCC

t

E

800

T

700TITITI

TI600 T,

TITI

5" 500

400

5300

200

100u

0 I I0 100 200 300

Time (s)

Figure 12. NEPTUN system response--testintermediate rod power.

200

otal test section - 180&p(mixture level)

level 53 level 4'Olevel6 - 160' level 3' level 7C level 8

level 2 140

12 0 E

C/ 0.100

o

,10

20

I 0

400 500

5001--low system pressure,

800

700

600

-7200

180

160

140

E

C

500

400

300

0

00-

120 E

c-0

100 -U0

8080 B

I.-

LA•

60

200

100

0

-440

20

0500

Time (s)

Figure 13 : NEPTUN system response - testpressure, high rod power, repeat

5009 - High systemtest

37

Wn,

Canmet bale 00,4

1 6

700

600

*500

4'00

E

300

100

//

-_____ __ / I/

| ! | ! f ! t !

0 100 200 300 400 500 600 700 am 900 100Tinme ($1

Figure 14.. Comparison of center rod Internal and LOFT theriocouples (testSO07. axial elevation. 946m -level 4).

amI

700

E•

z600

200

100

6 4 4 1

a i A a

- Centelr bae gOds

- LOFT thermocouple

•I I I) | 4m I I I I I I E I

0 100 200 300 400 50w 600 700 600 SOW 1000

Tie* (s)

Figure IS. ConParlson of center rod internal and LOFT thermocouples (testSO11. axial elevation. 946 an -level 4).

38

RELAPS/2/36.02 BOILOFF EXP. 50076.

S.

cc

A;

3.

2.

I.

9.

-1.

tem8.Title (SEC)

Figure 16: Water entrainment in NEPTUN boil-off experiment 5007, calculatedby RELAP5/MOD2, with and without new correlation for the inter-facial friction in bubbly and slug flow.

RELAPS/2/36.02 BOILOFF EXP. 5007

Z;

a;

1388.

9e8.

788.

S88.

388.a. 288. 488. 608. 8138. 2000.

Tilne (SEC)

Figure 17: Rod cladding temperature at measurement level 4 in NEPTUN boil-offexperiment 5007. calculated by RELAP5/tlOD2. with and without newcorrelation for the interfacial friction in bubbly and slug flow.

.&

2.292.t82.M8

I.90

1.78

I.,e

1.382.281.18

a.9"9.o0

8.788.68

8.58

0.300.28

L

7

sR

TRAC-BDI 4 ROD 17 CELL NEPTUN 5S02 (1) CHANGED HTCOR 2.282.132.083.981.881.78i.682 .58

1.3i1.28

2.88

0.908.888.788.60

8.488.308.20

.............................................

8. 280.88 AMU.8TIMEC 256CI

- TRX-903 M8OICTION EX0tR~tMATL. LM6E3

880.8

- TRAC-801 P'REDICTION - EXPER~rCNTAL LME3

1300.8W

9108.0a

Sea. as

M•0n

780. ea

788.88

308.8N

TRAC-BOI 4 ROD 17 CELL NEPTUN 5002 (1) CHANGED HTCOR TRAC-BDl 4 ROD 17 CELL NEPTUN 5002 (1) CHANGED FRCI.HTCOR

7

7

7 ..........

B

..................................... .211.88

900.00

388.88

988.18

888.8

B

! . .. . t , , , f . ...........

TR- -1C-0I CO•ONENT 2 RC0 GROUP I N1CE I2(2 . ROO 0 TA IRS-946

Fig. 18 : Comparison of measured and calculatedcollapsed liquid level and peak axialpower level rod surface temperaturehistories, using frozen version of TRAC-BDI for NEPTUN boil-off experiment 5002.

8. 208. 00 48.8 6W8.8 ea8.WTIME (SEC)

- TRAC-90 CGOM4( T 2 RMGRUPI OE 7 XP A (62M-1

Fig. 23: Comparison of measured and calculatedcollapsed liquid level and peak axialpower level rod surface temperaturehistories, using frozen version of TRAC-BDI for NEPTUN boil-off experiment 5002.

3888.80

TRAC-8D1 4 ROD 17 CELL NEPTUN 5006 (51 HTCOR

E

.J

,.J

2.272.182.991.901.80

1.791.691.50I.A2

1.391.291.101.00

0.699.59

0.40

0.390.29

A

•.........

...............

, ! . t . . . . . . ! . . . . . .* . . ... . . . .. . . .... . . . . . . .

2.2"2.122.091.90

1.981.78

:=I. W

R 1.581

2.69

18.49

2.39

2.29

TRAC-O0I A ROO 17 CELL NEPTUN 5006 (5) CHANGED FRCI.HTCOR.

... . A

. . . . , i , . . , . . . I . . . . . ! . . !

9. 59.e9 103.00 I25.9 29.99 259.9 ee 3 59.99 4M9.e 9 45..99 5ae.9e 550.eaTIME (SEC).__ TRAC-BOI PREDICTION ___ EXP"ERKr.NTAI. LEVEL

9. 59.M•9 2E9.9 I5.9 299.99 259.99 W.99 359.99 19.00 d5459.9 599.ee 59.P99TIMO 2I592

- TR9C-00 PREDICTION4 - 99769R6?471. LEVEL

11CO.C'o

I

TRAC-BDI A ROD 17 CELL NEPTUN 5006 (5) HTCOR

B.......... .......... ..........2299.99

2999.99

M9.9ea

g

B

............

7

TRAC-D01 A ROD 17 CELL,NEPTUN 5006 (5) CHANGED FRCI.HTCOR.

0

. . . . . . . . . . . . . . . . . , . , . . . . , . . . . . . . _ . . . . . . . . . . . . . .0. 59.90 290.00 IS29.9 2.9 25.0 3 i.; 359.99 e;9 . 15e99 Se9.90 559.99

TIME ($ECI

IR99-002 CMnONT 2 O CROUP 2I NOD .7... EXP. I=DA TA Tl9-946


9. 59.99 299.99 I2.99 29.99 255.99 3M9.08 35%.W 400.0. 459.99 5.UMTir2 I..C .

_ IR(X-001 CD? NT 2 ROO IMR•P I N•OE .7_.. EM POO DATA TftS-9j$

559.99

Fig. 24: Comparison of measured and calculatedcollapsed liquid level and peak axialpower level rod surface temperaturehistories, using frozen version of TRAC-BDl for NEPTUN boil-off experiment 5006.

TRAC-8D1 J ROD 17 CELL NEPTUN 5007 (5) CHANGED FRCI.HTCOR.2.292.182.99

1.901.e9

1.69C .59

1.39o 1.20

0. 0.

0.790.600.500.400.309.29

Me. ea

I 100.,ea

a

2.292.112.90

2.99

1.291.•9

1.59

I.49Mo39

1.29

t.18

0.999.6a9.79

9.59

9.49

9.39

9.29

........................

I. 59.69 159.69 259.69 25-,9.94e 459.e9 550.00 659.69 25.69 651.99 959.6WlrtE tsMCt I_ TRnC.-e0t I'MICTUOR E [VRIME•NTA LEVEL- TRAC-801 PREOICIION - EXPERIM¶ENTAL LEVEL

170A.P9

Hr:

9M. ee

B

See. s

- T169-801 COqIIIENT 2 Rt6 CS"r I OC EP. AM~ DATA 765-9AS- 56~t-69 COF~O412 6~95 ~ TIME (SEC) t 99 9.4



Ew

o

.J

2.20

2.20

2.001.98I."O1.8CC1.701.60

2.58

1.181.38

1.20

1.10I.ea

0.90

0.09

0.70

8.880.58

P.0

..J

2.202.102.801.901.801.701.1881.58

1.301.20

1.201.080.980.68

O.Se0.78

0.380.20

TRAC-B01 A ROD 17 CELL NEPTUN 5008 (5) CHANGED FRCI.HTCOR

.........................

TRAC-BOI A ROD 17 CELL NEPTUN 5008 (5) CHRNGED.HTCOR

1129.03

1208. ea

B

B

ROOMe2888.88

CEO.88o

w. M8.2.8 2 28e8o.ea

T- RC-801 PREOICTlO - E8XPRItENT81. LEVEi.

TRAC-80 A ROD 17 CELL NEPTUN 5008 (5) CHANGED FRCI.HTCOR

B '

700.8O

A8t.080. 18.8O 88O.88 12t 8.08 1 8.88 ZWO.80

12ME ISECI- TRAC-1301 DWI'08P 2 ROD CROUP I NODE 7t.... 888. 800 CAI RS1-948


0. MAO0 888 1208.08 1888.88ME8 ISECC

- TW-801 CCftWNNT 2 P00 ~P I NOD 7L.. EXP. POO Dl1A TRS-918

2808.08


s•

.J

2.202.102.6e1.901.08

1.782.681.50

2. 29r2.38

1.2"

0.98

6.880.780.60

.9500.406.380.28

,J

.J

2.202.16

2.981.9402.882.782.681.50

1.481.301.20

1.10

8.8419.708.606.98

6.166.380.20

TRAC-BOI 4 ROD 17 CELL NEPTUN 5011 (5) CHRNGED FRCI.HTCOR

. . A.

TIMlE 256CI- TRA-Bol P8EDICTIOR EXPESMIAI8 LiEL

280.68 258.83 3V

TRAC-BO1 4 ROD 17 CELL NEPTUN 5011 15)

JB

...... .. .. .. .. .

928.Ce

TRAC-BDI A ROD 17 CELL NEPTUN 5011 (5) CHPNGED FRC|.HTCOR

l.

.... .. ....LAJ

709.M8

6. 58.68 268.68 258.68 28.88TIME (SEC)

TInC-8OI COMPONNT 2 AM0 ~ I NO•I ,-.-....7 EXP. AM00 DATR TR-96

6. 58.6t 268.68 158.00 0088.6TIMNE 2iSTT

- 98-00-80 CO6CaNENT 2 AM0 CA01 I NME .2... MI. AM0 DATA TRS-946

258.68 30.600



NEPTUN 5006 (5) V12.VROD=O,NO SM.

C:

U

CC

1200.00

!100.00

900.20

700.20

600.e0

500. 0

400.00

.. ................... .... ./ .... '

"• ! !! I !!I! !

0. 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00 A50.00 500.00TIME ISEC)

- TR,.C-80I COMPONENT 2 ROD GROUP 1 NODE ,7-..-. EXP. ROD DRTA TRS-946

TRAC-BDI 4 ROD 17 CELL NEPTUN 5006 (5) HTCOR

Li

Li

LiI.-

Li(aCC

C0C

1200.00

1100.00

1000.00

900.00

800.00

700.00

600.00

500.00

400.00

0. 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00TIME (SEC)

-- TR~C-BD1 COMPONENT 2 ROD GROUP I NODE _....... EXP. ROD DATA TRS-946

Fig. 28: Rod surface temperature histories at peak axial power levelin NEPTUN boil-off experiment 5006 calculated by.TRAC-BDl,(a) frozen version, (b) modified for steam cooling.

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER(2-89) (Assigned by NRC. Add Vol., Sspp.o Rev.,NRCM 1102, and Addendum Number,, If any.)3201,3202 BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse)

2. TITLE AND SUBTITLE NUREG/IA-0040EIR-Bericht Nr. 629Boil-Off Experiments with the EIR-NEPTUN Facility:

Analysis and Code Assessment Overview Report 3. DATE REPORT PUBLISHEDMONTH YEAR

March 19924. FIN OR GRANT NUMBER

A46825. AUTHOR(S) 6. TYPE OF REPORT

S. N. Aksan, F. Stierli, G. Th. Analytis Technical7. PER IOD COVERED (Inclusive Dares)

B. PERFORMING ORGANIZATION - NAME AND ADDRESS 11f NRC, provide Division, Office or Region, U.S. Nuclear Regulatory, Commission, and mailing address,- if contra#ctor, providename and ma.,fing address.)

Laboratory for Thermal-Hydraul icsPaul Scherrer Institute (PSI)5232-Villigen PSISwi tzerl and

9. SPONSORING ORGANIZATION - NAME AND ADDR ESS (if NRC, type "Same as above"; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission,and mailing address.)


10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

A series of experiments was performed in the NEPTUN test facility consisting of tenboil-off (core uncovery) and one adiabatic heat-up tests. In these tests rod power,system pressure and initial coolant subcooling were varied. The repeatability of theexperiments was also demonstrated.

Some of the boil-off data obtained from the NEPTUN test facility are used for theassessment of the thermal-hydraulic transient computer codes. These calculationswere performed extensively using the frozen version of TRAC-BD1/MODI (version 22).A limited number of assessment calculations were also done with RELAP5/MOD2(version 36.02). In this report the main results and conclusions of these calculation!are presented with the identification of problem areas in relation to the modelsrelevant to boil-off phenomena.

12. KEY WORDS/DESCR!PTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT

Unl imitedICAP program, RELAP5/MOD2, TRAC-BDI/IIOD1, boil-off, NEPTUN 14.SECURITYCLASSIF ICATION

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Unclassified(This Report)

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