ML20003F878

Research Info Ltr 115:submits Results of Independent Assessment of TRAC-PIA Capabilities

ML20003F878
Person / Time
Issue date:	02/25/1981
From:	Minogue R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RIL-115, NUDOCS 8104230804
Download: ML20003F878 (15)

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NUCLEAR REGULATORY COMMISSION 5. (3,J(1)n i WASHINGTON, D. C. 20556

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February 25, 1981 i

MEMORANDUM FOR: Harold R. Denton, Director Office of Nuclear Reactor Regulation FROM:

Robert B. Minogue, Director Office of Nuclear Regulatory Research

SUBJECT:

- RESEARCri INFORMATION LETTER

1. 115 j.

INDEPElDENT ASSESSMENT OF TRAC-P1A COMPUTER CODE

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1.0 INTRODUCTION

The Research Information Letter (RIL) No. 92 dated June 1980 described the TRAC-PIA computer code for detailed, best estimate analyses of PWR LOCA. Following the release of this code to the public in mid 1978, RES centracted with three national laboratories to undertake an independent assessment of TRAC-PlA capabilities. Results of their efforts are sumarized herein.

The TRAC-PIA code was superseded by the TRAC-PD2 code version when the latter was released to the public in October 1980. Hence, although this infonnation letter describes strengths and weaknesses of the first version of TRAC code, the information here contained will be very valuable s

in determining whether substantial improvements were made in the subsequent code versions.

Independent assessment of the new version (TRAC-PD2) w;.

started in November 1980.

The overall assessment of TRAC will span three to four versions of that code as they are released to the pubite, starti.ng with TRAC-PIA. Results of hundreds of the integral, separate effects, and basic tests will be utilized over the next 4 to 5 years to quantify the code accuracy. Each new version of TRAC will be subjected to independent assessment utilizing, primarily, new portions of the code assessment matrix, with few repetitions involving those test cases which were poorly predicted with the previous version. RES plans to issue a research information letter describing results of independent assessment of each version of TRAC and RELAP-5 code. The attempt will be made, in each code assessment RIL, to identify the projected code accuracy for. LWR application. Reliability of that information will increase with each subsequent RIL as coverage of the assessment matrix increases. The overall code assessment matrix and-assessment methodology are described in Reference 1.

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FEB 2 51981 A

2.0 SUMFARY OF RESULTS The Brookhaven National Laboratory (BNL) provided the bulk of the assessment of TRAC-P1A physical models. As part of this effort, the BNL staff first surveyed the code documentation as well as FORTRAN listings to catalog all physical models, correlations, assumptions and data base.

Next, they compared code results against measurements from about 33

" basic tests" selected to illustrate strengths and weaknesses of the ado'pted models. In some instances, BNL performed sensitivity studies to examine consequences of certain model changes.

In addition, Los Alamos Scientific Laboratory (LASL) staff also perfonned comparisons of TRAC-PlA results against measurements from several basic

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tests. Listings of all basic tests utilized in assessment of TRAC-P1 A are shown in Table I.

Also shown are the laboratory descriptions (where

, the comparison was made) and overall remarks concerning the code's

. strong and weak points observed in each comparison.

A listing of about 25 Separate Effects Tests utilized in the assessment of TRAC-PIA is shewn in Table II. LASL was responsible for most of the comparisons made in this area.

The nine Integral Systems Tests used in TRAC-P1 A assessment are listeJ in Table Ill.

It can be seen that TRAC-P1A analysis of one of the LOFT Laboratory (INEL) performed by both LASL and the Idaho National Engineering tests (L2-3) was. This duplication was undertaken to explore the code user's effect on calculation results. As long as the code user has the liberty to select nodalization (discretization) of the system geometry and several input options, he will impact the code results in some measure; we wished to know how much. One of the major goals in the development of advanced cens was to minimize the user impact.

Finally, INEL was asked to " walk" the code through several challenging PWR LOCA scenarios which the code should be able to handle. They are listed in Table IV.

The detailed description and discussion of all calculations listed in Tables I through IV are given in References 2 through 11.

3.0 OBSERVATIONS (1) TRAC-PlA can calculate a complete LOCA analysis.

Instances of

" water packing" have been encountered on various occasions, requiring restarts of calculational segments utilizing smaller time steps.

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Harold R. Denton 3

FEB 2 51981 (2) The code is capable of performing multidimensional analysis of the reactor vessel thermal hydraulics. The results are intuitively correct. Quantification of accuracy of the multidimensional analysis will be made in the future (and with a future version of TRAC) when the required measurements from 2D/3D experiments become available. The code predicted the observed asymmetric refill through the LOFT downcomer. While BNL had difficulties in attempting to calculate the 2D test perfomed at the Ransselaer Polytechnic Institute (RPI), Sandia staff successfully predicted the jet expansion in the HDR test series.

(3) The six field equation treatment of the vessel hydraulics allows for consideration of tne thermal and mechanical non-equillibrium. The physical models dealing with the interfacial transport of mass, momentum, and energy are strongly dependent on the flow regime " map."

The latter is based on empirical observations pertinent to steady flows in vertically oriented pipes. While these simplifications appear to be adequate for the bubbly and the dispersed droplets flow regimes, current experience indicates that more sophisticated criteria are needed for the intemediate flow regimes. Such criteria may have to be specialized to different regions of the reactor vessel, to depict glo!.a1 phenomena.

(4) Further improvements are warranted in the models for non-equilibrium vapor generation and condensation, for liquid droplets entrainment and deposition, and for interfacial shear. Modeling of shear between the fluid and the wetted walls and, in particular, between the fluid and the embedded l

fuel rods, also needs improvement. Better empirical data base is, needed to generate an improved model for two-phase flow resistance offered by rod arrays when flows are not purely axial.

~ (5) Fuel rod quenching treatment in TRAC-PlA is superior to the reflood model in the most recent RELAP code version. Neverthe-g less, a need for a more mechanistic treatment of the quench front propagation was identified. These improvements were implemented in TRAC-PD2.

In addition, treatment of the critical heat flux in TRAC-Pl A does not differentiate between the departure from nucleate boiling (DNB) and the burnout or dryout.

l (6) Heat conduction within solids, other than the fuel rods, is calculated with a lumped parameter model. Finer radial or lateral discretization is necessary to handle hot wall effects which play an important role in the small-scale test facilities.

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FES 2 5151 (7) A one-dimensional, five-field equation drift flux model, with full themal non-equillibrium, is used in solving the transient two-phase flow within system components residing outside the reactor v' sel. This includes piping, pressurizer, steam generator, ECC accumulator, and valves.

i The difference between the vapor and the liquid velocities is prescribed through a constitutive relationship which is a function of the local flow regime.

Inadequate empirical knowledge exists for specification of this so called relative velocity for situations in which the -

vapor and the liquid flow in opposite directions. Such situations arise in certain small-break scenarios within the U-tube steam generacor primary tubing and within the hot leg. The code does not adequately handle the ficw regimes in horizontal piping when the fluid velocities are icw enough to cause phase separation.

Inability to handle phase separation in horizontal pipes and countercurrent ficw in one-dimensional flow paths greatly limits the code ability to handle those small-break LOCA scenarios where these processes play an important role.

(8) Treatment of the steam generator (S.G.) secondary side needs refinement in handling the liquid separation, its downflow and thermal mixing with the upcoming feed flow. More accurate tracking of the S.G. liquid level is also needed, and the ability to model the relief and the isolation valves.

Improvements are also required in the treatment of heat exchange between the primary and the secondary sides in the presence of l

vapor condensation.

(9) The critical flow through (pipe} breaks and through relief i i valves is handled in a mechanistic manner. This, however, requires very detailed spatial discretization near the break.

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Fully implicit numerical solution technique is employed for these regions to avoid the use of very small' time steps.

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While this procedure appears adequate for analyses of large and intemediate break sizes, it still presents a calculational burden which has a detrimental effect on the code runni.ng time I

for very small-break sizes. The code was not able to predict accurately the critical flow of high'.y subcocled liquid in some of the Marviken experiments. Ti.e BNL sensitivity studies indicate that the main cause is the mass exchange model which lj appears to exaggerate the local vapor production rate. BNL also recomends introduction of the nucleation delay model.

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Harold R. Denton 5

FEB 2 51981 (10) The code also suffers from inadequate mass conservation.

While its effect is not important in large-break LOCA analyses, it becomes very detrimental in small-break LOCA analyses.

Some numerical diffusion was also observed, as in every other code that employs the Eulerian fram of reference (e.g., RELAP5, COBRA,etc.).

(11) Improvements are needed in describing.he fuel behavior--

clad ballooning, gap conductance, thermal conductivity in fuel pellets, etc.

i CONCLUSION:

The TRAC-PlA code has shown a capability to address many and diverse transients involvi.ng both single and two-phase fluids. This was the i

first code capable of multidimensional treatment of the reactor vessel, within the context of the integral system analysis.

In~ the course of a rather extensive independent assessment of the publicly released version of that code, many weak poir.ts were indentified and communicated to the code developers. The de;elopment of the new version of TRAC (TRAC-PD2) has greatly benefited from these findings, and attempts were mar:e to remove as many of the identified weaknesses as possible. We are told that TRAC-P02 is a much more reliable, economical, and accurate code as compared to its predecessor. Validity of these i,

claims will be carefully examined in the course of TRAC-PD2 assessment j,.

currently undenvay.

Information on the predictive capabilities of the TRAC-PIA code reported in this Research Inforr;ation, Letter provides a reference for gauging the progress made with future versions of TRAC. Figure 1 illustrates the

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uncertainty in the TRAC-P1A prediction of the peak clad temperature (PCT) for the integral systems tests addressed in this assessment, as a function of the test facility scale. Extrapolation to full scale provides I

a very rough guidance r.egarding the uncertainty in the prediction of PCT for LOCA in PWR plants. Reliability of this type of information will improve as more cases are addressed in the assessment of future code versions.

TRAC-PlA represents the first version of an advanced systems code.

" Teething" problems were, therefore, fully expected and so is their removal, as the maturation process sets in.

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Harold R. Denton 6

FEB 2 51981 F

Nevertheless, the TRAC code was fcund to be not only adequate but also essential for the 1srge break LOCA analyses for PWR's, primarily due to strong multidimensional effects observed in some of'the cases listed in Table IV. At the same time, and for the reasons given under items (7) through (10) in the preceding section, the TRAC-P1A code is not recommended for application to analysis of the small break LOCA, or for other transients of long duration. For large break PWR LOCA best estimate analysis TRAC-Pl A is definitely superior to RELAr 4/ MOD 7.

D

%cbert B. Minogue, Director Office of Nuclear Regulatory Research -

Enclosures:

1. Table I-Basic Tests

2. Table II-Separate Effects Tests

3. Table III-Integral Systems Tests

4. Table IV-P'r."R LOCA Analyses

5. Figure 1-Diff. Betw. the Pred.

and the Meas._ Peak Clad Temp.

as Func. of Test Facil. Scale

6. References l

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TABLE I - BASIC TESTS Contractor Sunnary of Findings TEST CASE Performing Analysis

. Code Strengths

' Code Weaknesses i

I.

Critical & Subcritical i

Flows A.

Moby-Dick (Steady-State.

BNL Air-Water, Critical Flow)

1. Zero Quality Pressure drop Flow rate is sensitive to friction _..

(subcritical) factor option.

2. Low-intermediate Pressure drop - Void fraction Not a smooth steady-state and flow.

i Qualities (5 test conditions)

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3. High Quality No convergence to a steady-state.

B.

Moby-Dick (Steady-State, BNL No convergence to a steady-rate.

Steam-Water, critical flow)

C.

KFK - IRE BNL (Steady-State critical flow) 1.

Cold Water Accurate pressure drop and flow (subcritical) rate with homogeneous friction option.

2.

Air-Water Agreement of pressure flow rate and vapor fraction, with test data is below expectation.

3.

Steam-Water Vapor fraction calculation is good.

Prediction of pressure and flow rate below expectation.

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Tabl,e I - Page 2

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4 Contractor Sunraary of findings i

TEST CASE Performing Analysis Code Strengths Code Weaknesses D.

BNL Nozzle Tests

' BNL Good agreement of pressure, void Under-prediction of mass flow rate (Steady-State) friction and flow rates with that for subcooled inlet conditions.

measured at high flow rates.

Poor agreement with pressure at.10w 1 run subcritical flow rates.

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E.

Canon Test BNL Good prediction of pressure until Overprediction of vapor fraction.,

i transient, blowdown) 20 ms.

Poor predictions of pressures 4 test conditions) j after 20 ms.

F.

Super Canon Test BNL Reasonable prediction of vapor Poor prediction of pressure during the entire transient.

(transient, blowdown) fraction.

4 (4 test conditions) j G.

Edwards Test LASL Reasonable predictions of pressure Required additive friction factor.

(transient, blowdown) and temperature.

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H.

CISE TEST LASL j

(transient, blowdown with area changes, wall heat transfer, and gravitational effects) 1.

Unheated Good prediction of pressure, 4

temperature and mass inventory.

2.

Heated Prediction of pressure and temperatures below expectation.

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Analysis Code Strengths

' Code Weaknesses II. Counter-Current Flow Constitutive relationships do not A.

University of Houston BNL permit calculation of counter-current flow in a pipe component in churn-

' turbulent and annular flow regimes.

l Same comment as above. Possibly d'~s l

B.

Dartmouth LASL to inappropriate nodalization of the I

upper plenum.

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i-III. Multi-dimensional Steady-t State Tests I

I Code either failed to converge or A.

RPI, 2-D Phase BNL i

stopped, stating indefinite or j

Separation Tests i

overflow conditions.

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B.

FRIGG Rod Bundle BNL l

Code failed to converge.

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I TABLE II - SEPARATE EFFECTS TESTS Contractor Sui.uary of findings TEST CASE Perfonning Analysis Code Strengths Code Weaknesses I.

Marviken Tests BNL & LASL Reasonably successful predictions Underprediction of critical flow rates (full-scale Vessel Blow-of critical flow rates for i

for subcooled inlet conditions.

down) saturated inlet condithns.

This becomes more pronounced for nozzles with small length-to-(16 test conditions) l diameter ratio.

, II. Battelle-Frankfurt BNL & LASL Prediction of critical flow rate for_.

(Int.Std. Problem-steam and two-phase steam-water wer TcpBlowdown) not as good as expected. Possible i

experimental errors. Numerical l

j diffusion observed.

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(II. Semiscal 1-1/2 loop LASL llot leg break flow rate, system Prediction of cold leg break flow rate 6

Isothermal Blowdown pressure and temperatures are below expectation.

(Test 1011 Std. Prob. 2) well predicted.

j-i IV. Semiscale Hod-1 lleated LASL Reasonable predictions of pressure, Premature CllF.

i Loop Blowdown break flow rates and loop mass (Tests-02-8) flow rates.

V.

Crease Counter-Current LASL Good predictions of ECC bypass and Flow Test penetration using the vessel (Counter-currentflow, module.

4 subcases run) penetration, ECC bypass and VI. FLECilT Refloed Tests LASL Good predictions of clad Poor prediction of turnaround time, (forced-bottom-flooding, temperatures for high flooding quench time and quench temperatures reflood heat transfer, rate.

for low flooding rates.

quenching and liquid entra ttunent)

(3testconditions)

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TABLE III - INTEGRAL SYSTEMS TESTS Contractor Sunpary of Findings

• i, TEST CASE Performing Analysis Code Strengths

' Code Weaknesses I.

Semiscale Mod-3 LASL Reasonable agreement with test Poor agreement during refill and (Large Break LOCA Test, data during the blowdown.

and reflood, due to combination of Tests-07-6) code weakness in calculating down-comer penetration, reflood heat transfer, liquid entrainment and

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uncertainties in downcomer metal j

heat transfer.

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II, Semiscale Mod-l INEL Adequate prediction of clad Prediction of mass flows including l i '-

(Large Break LOCA Test, temperature when CHF was calculated break flow rates, pressures, l

densities, and fluid temperatures

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Tests-04-6) correctly.

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below expectation. Poor prediction S

of refill.

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II. LOBI Al-04 LASL Accurate prediction of peak cled Predictio'n of pressures, fluid

'j, (Blind, pre-testprediction, temperature. Accurate predictions temperatures and mass flow rates -

virgintestfacility) of CilF at different elevations.

below expectation.

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4 IV. LOFT L1-4 LASL Good predictions of mass flow i}!

(Isothermal blowdown with rates, fluid tunperatures,

'i' delayed ECC injection) densities, pressures and vessel liquid mass. Good representation of integral effects during the blowdown and refill phases.

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LOFT L1-5 LASL Good predictions of system Calculated liquid inventory in the (Isothermal blowdown) pressure and densities.

core is lower than that in the test.

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LOFT L2-2 LASL Good prediction of system pressure, Poor agreement in calculating rewet (LargebreakLOCATest, temperatures ECC behavior and in high powered region. However, 50% power) cladding temperatures at peripheral there is uncertainty in measurement rods including rewet behavior 4 of clad temperatures due to fin i

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.I Table III - Page 2 Contractor Sur.saary of findings

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TEST CASE Performing b

Analysis _

' Code Strengths

' Code Weaknesses effects which could also be LOFT L2-2 (continued) responsible for overprediction of peak clad temperature in high powered region. Underprediction of

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critical flow in subcooled inlet i

conditions.

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LOFT L2-3 LASL Reasonable prediction of system Same weaknesses as in L2-2 (Large break LOCA Test, pressure, temperatures, ECC 75% power) behavior and cladding temperatures at peripheral rods including rewet l

behavior.

LOFT L2-3 INEL Reasonably good agreement with Failed to predict positive core inlet 4

(Large Break LOCA Test, experimental data early in the flow early during the LOCA and i

75% power) transient.

consequently failed to predict rewet.

Predictions of break flow rates, I

pressures, pressure differentials.

densities, and clad temperatures below expectation.

Inaccurate calculation of mass i

LOFT L3-0 LASL conservation. Accuracy of predictions (Isothermal blowdown of all parameters also suffered from throughPORV) uncertainties in internal vessel leakage (downcomer to upper plenum; and in additive system leakage beyond fl ow.

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Same weaknesses as in L3-0 LOFT L3-1 LASL f

(Small Break LOCA Test, 100% power) j e

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TABLE IV - PWR LOCA ANLAYSES i

Qualitative Assessment of. Code Ferfonnance dnd of Results LOCA Scenario

.1.

Large Cold Leg Break Satisfactory and reasonable, j

e (200%)

l 2.

Intermediate Cold Leg Satisfactory and reasonable.

l Break (0.25m-diameter) 4 3.

Small Cold Leg Break Results not reasonable.

(0.10m-diameter)

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4.

Large Cold Leg Break Satisfactory and reasonable.

important 3-D effects observed in results, i

with rupture of steam l

generator tube (s) j (Twocases,onewith small and the other s.

with large number of tubes) 5.

Large llot Leg Break Satisfactory and reasonable.

(200%)

i 6.

Large llot Leg Break Satisfactory and reasonable, j

i with rupture of 16 steam generator tubes l

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1 REFERENCES 1.

S. Fabic and P. S. Andersen, " Plans For Assessment of Best Estimate LWR Systems Codes," NUREG-0676, January 1981.

2.

P. Saha et al., " Annual Report On TRAC Independent Assessment at BNL,"

BNL-NUREG-27580, January 1980.

3.

U.S. Rohatgi, P. Saha, " Constitutive Relations in TRAC-P1 A,"

BNL-NUREG-51258 NUREG/CR-1651, August 1980.

{

4.

S.V. Lekach, " Calculation of the CANON Experiment Using the TRAC Code,"

{

BNL-NUREG-28290, August 1980.

T. Knight, " TRAC-P1 A Assessment - 1979," LA-8477-MS, NUREG/CR-1652 5.

(to be issued, copy available in Branch).

6.

A.C. Peterson, " TRAC-P1 A Independent Assessment Sumary Report,"_ __._

EGG-CAAP-5147, April 1980.

7.

P.D. Wheatley, " Comparison of TRAC-Pl A Calculations With LOFT L2-3 Experimental Results," EGG-CAAP-5072. December 1979.

p; 8.

P."

Demie, "An Analysis of Semiscale M00-1 LOCE S-04-6 Using the i

TRA -P1A Computer Program," EGG-CAAP-5181, June 1980.

5 9.

J.R. Larson, " Calculations of a Large Cold leg Break With Steam Generator Tube Ruptures In a PWR Using the TRAC-P1A Computer Program,"

'~

t f

EGG-CAAP-5189, June 1980.

10.

P.D. Wheatley, M.A. Bolander, " TRAC-P1 A Calculations for a 200%,

0.25m-Diameter, and 0.10m-Diameter Cold Leg Break In a Pressurized Water Reactor," EGG-CAAP-5190, June 1980.

M.A. Bolander, " TRAC-PlA Calculations for a 200% Hot leg Break and a 200% Hot Leg Break Simultaneous With a Rupture of 16 Steam Generator Tubes In a Pressurized Water Reactor," EGG-CAAP-5191, June 1980.

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