ML19212A537
| ML19212A537 | |
| Person / Time | |
|---|---|
| Issue date: | 06/07/1978 |
| From: | Levine S NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Case E, Minogue R Office of Nuclear Reactor Regulation, NRC OFFICE OF STANDARDS DEVELOPMENT |
| Shared Package | |
| ML19211A051 | List: |
| References | |
| RIL-029, RIL-29, NUDOCS 7912140502 | |
| Download: ML19212A537 (16) | |
Text
NRC 'Ph M %
- f UNITED FIATES NUCLEAR REGULATORY COMMISSION 3
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JUN 71978
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Z MEMORANDUM FOR:
E. G. Case, Acting Director
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Office of Nuclear Reactor Regulation
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P,. B. Minogue, Director l
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Saul Levine, Director J
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Office of Nuclear Regulatory Research f
SUBJECT:
RESEARCH INFORMATION LETTER
=29 FUEL R0D ANALYSIS COMPUTER CODE: FRAP-T3 Introduction and Sumnary This memorandum transmits the results of completed research to prepare and test the third modification of the computer code FRAP-T (Fuel Rod Analysis Program - Transient).
FRAP-T is a FORTRAN IV computer code being developed to predict the transient response of a LWR fuel rod during postulated accidents such as Loss-of-Coolant Accidents (LOCA),
Power Cooling Mismatch Accidents (PCM), Reactivity Initiated Accidents (RIA), or Inlet Flow Blockage Accidents (IFB).
FRAP-T is also being developed to perform the calculations needed for planning and analyzing Power Burst Facility (PBF) and Loss of Fluid Test (LOFT) experiments.
Although the code calculations are made on a best estimate (BE) basis, substitution of alternate models and correlations could be easily made to make evaluation model (EM) calculations.
FRAP-T3 is the third annual update of.the code and as such provides a relatively mature analytical capability.
Improvements upon FRAP-T2 are primarily in the area of cladding behavior.
Aspects of various versions of the code are shown in Table I.
The importance of improving our fuel behavior codes is recognized in a series of user requests:
REG:RSR-88, " Fuel Pin Analysis Development,"
dated March 14, 1973; REG:RSR-ll8, " Regulatory Need for Additional Safety Research on Reactivity Initiated Accidents," dated November 21, 1973; Section 6.8 of the " Regulatory Assessment of the AEC Water Reactor Safety Research Program," dated August 12,1974; " Review of Fuel Behavior Project Description," dated May 6,1975; "NRC/NRR Technical Safety Activities Report," dated September 11, 1975.
's219030s k
1570 240 m
25'.L s
E. G. Case R. B. Minogue D
These user requests are for analytical models, tested against data, which will predict fuel failure and failure propagation thresholds in power reactors. A calculational tool is also needed to interpret PBF, LOFT and Halden experiments, to provide audit capability for vendor codes such as STRIKIN-II, FACTRAN, LOCTA-IV and THETA, and to support specific SD and NRR activities. This memorandum and its enclosures describe the FRAP-T3 code, its testing and our evaluation of its applicability and capability.
Results - Code Features In FRAP-T3 the coupled effects of mechanical, thermal, internal gas and materiai property response on the behavior of the fuel rod are considered.
Given appropriate coolant condition and power histories, FRAP-T3 can calculate rod behavior for a wide variety of off-nonnal situations and postulated accident conditions (e.g., BWP, or PWR power transients, flow coastdown, load loss or coolant depressurization).
Further details of code features (e.g., models, input requirements, output parameters) are given in Appendix B.
Results - Code Qualification _
An essential part of producing an operational computer code, which can be used with a known degree of confidence, is the independent testing process (described on pages 257-267 of Appendix C).
The results of such testing of FRAP-T3 are as follows.
Figure 1 compares measured and pre-dicted centerline fuel temperatures for unpressurized rods.
The good agreement, generally within 101, suggests that heat transfer is well represented by the MacDonald-Broughton (" cracked pellet") gap conductance option which was used for these calculations.
Figure 2 indicates a similar comparison for rods prepressurized with helium, showing less satisfactory agreement.
However, a second FRAP-T3 gap conductance model is available, following Ross-Stoute, and this option provides good thermal predictions for prepressurized rods.I Figure 3 shows predictions of plenum gas pressure. Most of the high pressure results fall within 10%
of the measured values. Accurate prediction of this pressure is important to the ballooning behavior of fuel rods in a hypothetical LOCA.
Figure 4 compares single rod PBF (annulus geometry) test data with calculations using two of the Critical Heat Flux (CHF) options which are available to FRAP-T3. Lack of a better fit may be accounted for by peculiarities of the PBF test train configuration (e.g, standoff screws and flow area).
Figure 5 shows fuel temperature response following scram.
An adequate TFBP-TR-186, " Gap Conductance Test Series, Test GCl-3, Test Results Report I
and Summary of Piggyback Tests," March 1977.
1570 241
o E
E'. G. Case R. B. Minogue calculation of the dissipation of stored energy and decay heat immediately af ter. scram is especially important for analyzing accident situations.
Finally, Figure 6 compares results of a standard problem run with FRAP-T An agreement between these two independent
&nd with the German code SSYST.
codes implies some validity of the code predictions.
Evaluation In the context of LWR system transients, FRAP is well suited to be used as a component code to describe fine details of fuel rod behavior.
Furthermore, sensitivity studies with FRAP will facilitate definition of Substantial the simplest acceptable fuel description in systems codes.
effort has gone into FRAP-T3 to make it a mechanistic and sophisticated The independent testing process has shown that several raw models code.
and subcodes, some of which are unique to FRAP-T, are important to making 2
These include the material properties package,
realistic calculations.
the failure subcode, three dimensional cladding ballooning, a complete 3
heat transfer correlation package, a transient plenum temperature model The material properties package and the and an axial gas flow model.
failure subcode have been well received by the Fuel Code Review Group.
Quantitative characterization of the uncertainty associated with parameters predicted by FRAP-T3 (e.g., plenum pressure, fuel centerline temperature, cladding ballooning or burnout power) has been made, and representative samples are shown in the figures.
Developments are continuously underway to remove some of the present limitations of applicability of FRAP-T3 and these developments will be incorporated in future versions of FRAP as new research data and modeling permits.
FRAP-T3 has been transmitted to the Argonne Code Center and is programmed and running on the CDC 7600 computers at INEL (Idaho), Berkeley (California)
We would be happy to assist your staff in and Brookhaven (hew York).
running any of the FRAP standard problems listed in Table II in order to directly demonstrate the code's capability.
TREE-NUREG-1005, "A Handbook of Materials Properties for Use in the 2
Analysis of Light Water Reactor Fuel Rod Behavior," December 1976.
3 FBP-TR-189, " FRAIL 3: A Fuel Rod Failure Subcode," April 1977.
T 1570 242
q-E. G. Case R. B. Minogue Appendices Appendix A contains the six figures and two tables referred to in the Appendix B provides a succinct description of code features and text.
some coments concerning use of the code.
Appendix C, report TFBP-TR-194, "FRAP-T3-A Computer Code for the Transient Analysis of Oxide Fuel Rods," provides detailed descriptions of the code afid its testing.
1 s
Saul Levine, Director Office of Nuclear Regulatory Research
Enclosures:
As stated 1570 243
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APPENDIX A 1570 244
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Symboi Ru 2
MPR 80 2000 -
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IF A 431 p
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2000 0
500 1000 1500 Measured Fuel Centerkne Temperature (*C)
Predicted versus measured fuel centerline temperature for unpressurized rods.
FIGURE 1
1570 245
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2500 1
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D Symbol Run Reference 273'lPBF 274) 276 PBF 277 2000-PBF 278 V
g 280 PBF IF A-429 0
s To 430 PBF e
433
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1 0 0 500 1000 1500 2000 Measured Fuel Centerline Temperature ('C)
Predicted versus neasured fuel centerline temperature for pressurized rods.
FIGURE 2 1570 246
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9 439 200 oh 0 >- - I - - --- - 3 I
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O 200 400 600 800 1000 1200 1400 1600 1800 Measured Rod Internat Pressu e (psial r
Predicted versus measured rod internal pressure during heatup.
FIGURE 3 1570 247
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Run10 0014 B015 8016 Reference PBF ii FRAP.T3 Predictiona JCHF Correlat.or O
B&W 2 10 y_ - _7 3
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System Pressu e 2160 2230 ps.a r
O8 inlet Entnaicy 644 Btu io I
meatec Dia Hycraviic Dia 25-30 c0 inlet Quaiity 0 a
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06 g
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03 O 05 0 10 0 15 0 20 0 25 0 30 0 35 0 40 0 45 Thermal Load on Subchannel (kW >cm))
PBF Inlet mass flux at CHF onset vs rod themal output for low flow area tests.
1570 248 FIGURE 4
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-- - - FRAP-T3 Prediction.
. Halden data 4 ~ Predicted power adjusted to match initial measured temperature o,am Gap 2 5-2 7%
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12 16 20 24 28 32 36 40 4
Time after Scram initiation (sec)
FRAP-T3 fuel temperature response after scram at 5600 mwd /t.
1 FIGURE 5
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A Experimental Data FRAP-T T
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10 20 30 40 50 60-70 80 90 TIME (s)
FRAP-T3 and SSYST Predictions for Standard Problem PNS-4238 LOCA Heatup Test rv LD CD FIGURE 6
E TABLE I C AP ABILITIES OF V ARIOUS VERSIONS OF FRAP-T Phenomenon FRAP-T2 FRAP-T3 Modeled FRAp-Tl Heat conduction Stacked 1-0 radial Stacked 1-D radial Stacked 1-0 radial, 2-D r-e Modified Ross and Stoute Modified Ross and Stoute, Modified Ross and Stoute, Gap conductance Cracked pellet Cracked pellet Plenum gas temperature Coolant temperature Six-node transient Six-node transient
+ 10 F energy balance, boun-energy balance, dary conditions from simplified boundary surface temperature conditions subcode Metal-water reaction Baker-Just Baker-Just Cathcart Internal pressure Compressible, laminar Compressible, laminar Ideal gas law, gas flow, constant gas flow, constant compressible, laminar Hagen number Hagen number gas flow, variable Hagen number open porosity considered Cladding defonnation Uncoupled stress-strain Triaxial coupled plastic Triaxial coupled plastic stress-strain equations, stress-strain equations, equations, no fuel-cladding inter-fuel-cladding inter-fuel-cladding inter-action, no ballooning action, intermediate action, advanced w
N balloon model, no balloon model, strain-CD model, no creep rate effects, cold-creep work and fast neutron N
flux effects, computation
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optimization, no creep.
No model No model ANS model 5.1 (1971) t.
Decay heat s
B
.E TABLE I (continuedl Phenomenon FRAP-Tl FRAP-T2 FRAP-T3 Cladding failure failure if instability failure if total Failure probability
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strain exceeded circumferential strain computed, overstress, exceeded overstrain, eutec ic melting, and oxida(tion failure types modeled Fuel deformation GAPCON-1 Model GAPCON-I Model, GAPCON-l Model, free thermal free thermal expansion expansion model model High flow film boiling Groeneveld Groeneveld Groeneveld heat transfer Douga11-Rohsenow Dougal1-Rohsenow correlations Tong-Young Tong-Young Condie-Bengston Condie-Bengston low flow film Berenson Groeneveld Modified Bromley (a<0.6) boiling heat free convection (a1.G) 0 transfer correlations Critical heat flux B&W-2 B&W-2 B&W-2 correlations Barnett W-3 W-3 Modified Barnett Barnett Barnett 7(
Modified Barnett Modified Barnett General Electric General Electric CD Slip ratio correlation Homogeneous Modified Bankoff-Marchaterre-Hoglund N
Jones tn N
Water properties RELAP3 tables Wagner steam tables Wagner steam tables Fuel, cladding and MATPRO-2 MATPRO-6 MATPRO-8 gas properties O
g n
W TABLE 11
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FRAP-T STANDARD PROBLEMS TYPE DESCRIPTION DATE LOCA PWR Cold Leg Break Using Supplied Heat Transfer Coef ficients or RELAP Coolant Conditions LOCA TRE AT Test 2, BWR Rods 1971 Slow Power Ramp Halden Reactor Project, Norway 1967 Power-Cooling Mismatch PBF Test PCM 8-1 1976 Reactivity initiated Accident (RI A)
BWR Hot Standby Conditions,250 Cal /G g
u, RlA SPERT Test GEX-692 1969 Nu, ATWS BWR Main Steam Isolation Closure Valve Accident (90% Relief) s i
s
APPENDIX B l
f DESCRIPTION OF CODE FEATURES The phenomena modeled by the code include:
(1) heat conduction; (2) elastic plastic cladding deformation; (3) fuel-cladding mechanical inter-action; (4) transient fuel rod gas pressure; (5) heat transfer between fuel and cladding; (6) cladding-water chemical reaction; and (7) heat transfer from cladding to coolant.
Consideration of the mechanical deformation of the fuel and cladding is of particular significance, since a realistic prediction of rod geometry during an accident (e.g., LOCA) is desired.
The probability of pellet cladding interaction related failures is calculated, even though the models needed for a true description of local effects are missing.
Effects of prior irradiation must be input from another source (e.g., FRAP-5).
FRAP-T3 is linked to a modular material properties package, MATPRO-8, which contains correlations for all fuel, cladding, and gap gas properties needed by the code.
Each correlation is contained in a separate function subprogram or subroutine.
No material properties need be specified by the code user.
FRAP-T3 is also linked to the Wagner water properties package, which was developed for the RELAP-4 code.
This package defines subcooled, saturated, and superheated water properties.
FRAP-T3 requires input data (in either metric or engineering units) which specify cold state fuel rod geometry, transient power, transient condition of coolant surrounding fuel rod, and amount (or pressure) and type of gas in the fuel rod.
Input data are also required to specify mesh size (radial and axial incremental dimensions used in computation), time step and accuracy.
This permits the code user to have some control over the computer CPU time needed to execute a problem.
Transient coolant condi-tions can be specified in several ways.
These options have been chosen to provide maximum flexibility.
For example, card input of coolant condi-tions or heat transfer coefficients, or magnetic tape input of coolant conditions calculated by RELAP-4 can be used.
Code printout, which occurs at input specified time intervals, includes fuel rod radial temperature distribution at an arbitrary number of axial positions, fuel diameter, gas gap thickness, gap conductance, cladding diameter, axial length change, internal pressure, power, surface heat flux, and cladding hoop strain.
The code can be instructed to generate plots of the above output parameters as a function of time.
It is also possible to generate 16mm motion pictures of the output.
1570 254
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2-Based.on ouc review of FRAP-T3, we believe the following observations would be helpful to code users; (1) At hot plenum pressures above 500 psi the MacDonald-Broughton gap conductance model (so callea cracked pellet model) predicts excessive values and the Ross-Stoute model option is recommended.
(2) Two model options are available for computing fuel radial displacement (free thermal expansion model or GAPCON-THERMAL-I model).
Since FRAP-T3 was verified (and to some extent developed) using free thermal expansion, that model option is recommended.
(3) The stress-strain model in MATPRO is not applicable above 1500 F (temperature at which a metallurgical phase transformation begins in Zircaloy).
This generally causes an overprediction of cladding circumferential strain at burst.
Measured strains of 0.1 to 0.7 in/in are predicted to be 0.6 to 0.9 in/in.
O 1570 255
-. - - _ _