ML20003F931
| ML20003F931 | |
| Person / Time | |
|---|---|
| Issue date: | 01/06/1981 |
| From: | Minogue R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RIL-109, NUDOCS 8104230905 | |
| Download: ML20003F931 (17) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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6S MEMORANDUM FOR:
Harold Denton, Director Office of Nuclear Reactor Regulation FROM:
Robert B. Minogue, Director Of fice of Nuclear Regulatory Research
SUBJECT:
RESEARCM INFORMATION LETTER
- 109 WRAP-BWR-EM (WATER REACTOR ANALYSIS PACKAGE - BOILING WATER REACTOR - EVALUATION MODEL) 1.
INTRODUCTION:
OVERVIEWOFWRAP-BWR-EMDENLOPMENT The WRAP-BWR-EM computer code system (Ref. 1) is designed to provide NRC with the capability to perform audit calculations for Loss of Coolant Accidents (LOCAs) in boiling water reactors.
This code system is an cutgrowth of, and replacement for, the original WREM system (Ref. 2) of Evaluation-Model-Audit codes.
The work was performed in response to a request (Ref. 3) frem the Office of Nuclear Reactor Regulation to provide a complete Evaluation Model (EM) code package for' audit capability.
A companion Research Information Letter (RIL) is being issued for the WRAP-PWR-EM system for pressurized water reactors.
The main thrust of the system development was to use existing computer codes, each code calculating a specific facet or portion of a BWR LOCA, and provide automatic data transfer and interfacing between those codes.
Doing this allows the calculation to proceed smoothly through a complete LOCA sequence.
Considerable effort also went into making the system
)
format user convenient (Ref. 4) and also providing an automatic initializa-l tion for the LOCA transient (Ref. 5).
I It is worth mentioning that the Office of Nuclear Regulatory Research (RES) management of the WRAP program at the Savannah River Laboratory (SRL) was aided by the successful interaction between SRL and other groups of workers.
First, NRR personnel aided in the WRAP program from the beginning.
l Although they could not respond in writing to all requests for review of l
models and interfaces, their contributions were helpful (Ref. 6).
- Second, the check-out and verification of the completed WRAP-PWR-EM system was performed in parallel efforts by both Savannah River Laboratory (SRL) and the Idaho National Engineering Laboratory (INEL) (Ref. 7).
The assistance that INEL provided to SRL throughout this project was very important.
Third, the RISO Institute in Denmark at which one of the codes in the package was developed, provided assistance in resolving various problems associated with the installation of NORC00L at SRL.
I 8104230
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II.
DESCRIPTION OF CALCULATION SCHEME A.
Codes Forming The WRAP Package The WRAP system for BWR-EM analysis comprises several computer codes which have teen developed to analyze individual phases of a LOCA.
These codes include GAPCON (Ref. 8) for calculation of initial fuel conditions, TWRAM (the SRL analog of RELAP4/M005 (Ref. 9)) for analysis of the system blowdown, NORC00L (Ref. 10) for analysis of the reflood phase, and MOXY (Ref. 11) for the calculation of the fuel assembly hot plane temperature.
The GAPCCN module calculates preaccident thermal conditions of the fuel rods.
For a given fuel rod, GAPCON determines the gap conduc-tance, temperatures, pressures, and stored energy as a function of the power history of the rod.
These data are then used as initial conditions for tne transient fuel models.
GAPCON calculations can be performed as part of the WRAP-EM r-
'ar path or separately from the other calculations.
In either pro aure, GAPCON results may be storea in a cata library.
The stored GAPCON d:ta are then used as input to WRAP and MOXY calculations.
The GAPCON data transferred to WRAP require only two fuel rod cescrip-tions--an average rod for the hottest fuel bunale and an average rod for the otner fuel bundles.
The BWR system blowdown is calculated by the well-known RELAP4/M005 code (Ref. 9).
Tiie initial WRAP system was based on RELAP4/M005/
Version 65.
The initial step in the EM devel0pment required updating WRAP to RELAP4/M005/ Version 74 and then implementing several modi-fications to provioe an EM treatment of the BWR blowdown calculation.
The latter included:
Vertical slip modeling modifications (Ref. 12) necessary to o
properly model gravity-induced velocity differences between liquid and gas phases, Jet pump modeling modifications (Ref. 12) required to eliminate o
the discontinuity in the jet pump momentum equations that develops as either drive or suction flow becomes zero, Addition of the GEXL correlation (Ref.13) for determination of o
the boiling transition location, and Corrections to the fuel rod plenum temperature calculations o
(Ref. 12).
Several other corrections (Ref. 14) were made which included the proper calculation of potential energy contributions to the ju.xtion enthalpy when flow rev3rses as well as other minor coding modi-fications.
The M0XY module is used to perform thermal analysis of a planar section of a BWR fuel assembly during a LOCA.
The module calculates the temperature distribution of each fuel rod in a BWR fuel assembly at a single axial level of the assembly, usually the level with the hottest axial temperature. The code models descr.ibe heat transfer by conduction, convection, and radiation and heat generation by fission product decay and the metal-water reaction.
Fuel-rod swell-ing and rupture are considered along with energy transfer across the fuel-cladding gap.
Heat transfer in BWR assemblies is calculated during the three stages of a LOCA; blowdown, core heatup, and emergency cooling. Only during blowdown are the time-dependent data for power, heat transfer coefficient, and coolant temperature required.
For the remainder of the transient, built-in data as required by 10 CFR 50, Appendix K (Ref.15) guidelines for EM analysis are used.
The NORC00L code (Ref. 10) was developed for tha evaluation of emergency core cooling systems in BWRs.
It is applicable to the regime following the termination of reactor blowdown where there is near pressure equilibrium between the reactor vessel and the contain-ment.
The code is used to analyze operation of the spray and reflood systems through the refill and reflood phases.
NORC00L consists of two basic mooels:
a fuel-rod model ano a model for two phase flow. One-dimensional heat conduction equations are solved by the fuel rod model.
The two phase flow model is based on a solution of the multifield conservation equations for mass, momentum, and energy as well as the equation of state.
The flow regimes covered are single phase liquid, bubbly flow, inverse annular flow, film flow, and dispersed flow.
Thermodynamic equi-librium is not assumed; i.e., steam is allowed to be superheated and water subcooled.
The fuel-rod model and the two phase flow models are coupled through a number of physical models and heat transfer correlations which include conduction, convection, and thermal radiation effects.
Two-dimensional axial conduction is treated in the rewetting front.
B.
Code Interfaces For Data Transfer The integration of the GAPCON, RELAP, MOXY, and NORC00L modules to form the WRAP-EM system required the defining and programming of the following interfaces:
o GAPCON/RELAP, o
GAPCON/MOXY, o
RELAP/MOXY, 0
RELAP/NORC00L, and o
NORC00L/MOXY.
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These interfaces automate the computational steps required to perform a complete LOCA analysis from break through reflood.
In general, hot assembly and average assembly fuel pin conditions as calculated by GAPCON are transferred to WRAP via the GAPCON/RELAP interface.
The fuel pin conditions for the hot assembly are also transferred to M0XY via the GAPCON/MOXY interface.
The RELAP/
NORC00L interface is a transfer of the data from RELAP at end of blowdown to NORC00L for initialization of the reflood calculation.
Transient hot assembly data during blowdown and reflood are passed to M0XY via the RELAP/MOXY and NORCOOL/MOXY interfaces, respec-tively.
A more detailed description of the data transferred by each interface is presented in Appendix I and Reference 4.
C.
BWR Steady-State Procedure The RELAP4/M005 code provides no explicit procedure for initiali:ing the transient thermal-hydraulic calculation.
Instead, the user is required to generate the initial-system state by a series of nand calculations to produce estimates of the state variables and then short transient runs to evaluate the reasonableness of the estimates.
In the extension of the system for BWR-EM analysis, an automatic steady-state procedure has been developed for boiling water reactors.
The BWR steady-state procedure employs a time-decendent solution of the thermal-hydraulic equations without perturbation.
Instead of specifying the volume variables and junction flows required by RELAP, the WRAP user specifies the following quantities:
o Core power, o
Total water mass in the system, o
Steam dome specific volume, and o
Feedwater junction specific enthalpies.
~
The procedure then computes the:
o Thermodynamic state of all control volumes, and o
Flow rates for all junctions including the feedwater and steam line.
The uniqueness of the solution obtained by the BWR steady-state procedure has been demonstrated. The procedure requires approxi-mately 15 to 20 minutes computer (cpu) time on an IBM 360 Model 195 for a typical BWR nodalization. A detailed description of the procedure is provided in Reference 5.
D.
WRAP-BWR-EM Analysis Sequence The BWR-EM analysis scheme is shown in Figure 1.
The calculation proceeds along two paths, one for thermal / hydraulic analysis and the other for fuel rod response analysis.
This scheme has been reviewed by NRR personnel (Ref. 16).
At selected times during the calculation, information is' automatically passed from the thermal / hydraulic analysis t3 the fuel rod response analysis.
These include fuel thermal conditions for the hot plane analysis by the P0XY module and the times at which the following events occur during blowdown:
1.
Time of end of lower plenum flashing, 2.
Time when core sprays are at rated flow, 3.
Time for end of critical flow at the break.
The above times ao c e d in applying the 10 CFR 50, appendix K rule (Ref.15) for specifying convective heat transfer coefficients for BWR fuel unde scrs. cooling.
At the e o of blowdcon, system renodalization is performed by the RELAP/NORC00L interface routine and the reflood phase of the accident is calculated by the NORC00L module.
The time to hot plane quench is passed to MOXY for use in determining tne end time for the hot-plane analysis.
III. WRAP-BWR-EM VERIFICATION RESULTS Calculations to verify the capability of the WRAP-BWR-EM system were performed and compared with test data as well as with vendor calcula-tions.
All calculations, except the special M0XY/GE comparison, were performed at SRL with automatic interfacing between the separate codes, and with the same code versions at INEL (Ref. 7) with hand interfacing between codes. The latter procedure was used to uncover and correct coding and input errors.
M0XY calculations were compared to a General Electric (GE) EM analysis for a BWR/6 LOCA (Ref. 17).
Data supplied by NRC for a BWR/6-218 reference study were used for the MOXY calculations.
Data used for the GE calculations were not completely specified (Ref. 17).
Nevertheless, there was good general agreement between the MOXY results and the GE results. MOXY predicted essentially the same time-dependent behavior for the peak clad temperature (PCT) as did the GE methods.
For the input values used in this study, all values of PCT predicted by MOXY were within 7 percent of the GE value.
However, sensitivity studies performed with M0XY showed variations in PCT as large as 200*F, depending on values of input data which were estimated from plots in the GE report.
There was general agreement with the GE results for the time-dependent behavior of the average fuel temperature and the fuel pin pressure.
l However, MOXY predicted much higher initial average fuel temperatures and u-**
fuel plenum gas pressures than reported in the GE work.
This was probably due to differences in the initial fuel conditions such as power levels and distributions, cald gas pressure, and time in reactor.
The objective of this study was nct so much to exactly duplicate the GE results, but to demonstrate that MOXY could calculate the same general fuel-temperature behavior; this demonstration was achieved.
As part of the checkout program of the WRAP-BWR-EM system, companion calculations were conducted at INEL and SRL for a pump suction line break in a BWR/4-type plant (Ref. 18).
The Hope Creek plant was used for the reactor model.
RELAP4/M005 and M0XY input were created at INEL.
These input cards were converted to JOSHUA records suitable for use by the WRAP system of codes at SRL.
The results of the SRL blowdown calculations with the WRAP module TWRAM were in excellent agreement with the reported RELAP4/M005 blowdown calculations conducted at INEL.
Good agreement also was obtained between SRL-MOXY and INEL-MOXY calculations of the fuel heatup following the beginning of lower plenum flashing.
A verification calculation was also performed to cemonstrate the applica-tion of the package to the analysis of a loss-of-coolant accident for a General Electric BWR/6-218 design (Ref. 19).
The proDiem selected by NRC was a double-ended guillotine break of the pump suction line of a recircu-lation loop.
This same calculation is the base case large break fer a BWR/6 EM reference study currently being su ported at SRL by NRR.
The WRAP results were compared to an analogous GE calculation and the ongoing NRR study will perform parametric calculations to investigate the differencu in the WRAP and GE calculations.
The nodalization of the reactor system used in this study (Ref. 19), as well as many of the model options employad were originally developed at INEL as part of a best-estimate study.
Appropriate modifications were made to create an EM problem.
Representative fuel data were obtained from NRC (See Ref. 19).
All computational modules in the WRAP-BWR-EM system were exercised.
A steady state was not computed using the BWR steady-state module in WRAP since INEL had already established steady-state input parameters.
However, a supplementary calculation with WRAP yielded a steady-state very similar to the INEL input.
The reasonable comparison between the WRAP and GE calculated pressures is shown in Figure 2.
However, the clad temperature histories in the WRAP and GE calculations were quite different, as shown in Figure 3.
GE calculated departure from nucleate boiling (DNB) at the high powered plane at 10 sec; WRAP did not calculate DNB at that location until 48 sec.
The cause is probably a combination of two effects:
- 1) the predicted core flow during the first 10 sec of the transient, and 2) the use by GE of the GEXL correlation which is not applicable to low or reverse flow situations. Mnre detail is given in Appendix II.
As mentioned above, an ongoing parametric study is investigating these discrepancies, as part of the audit function of WRAP-BWR-EM.
Finally, WRAP-BWR-EM calculations were compared with data from two TLTA t
tests:
6007 (no ECC) and 6406 (with ECC), as presented in Reference 20.
These two tests were chosen for simulation with WRAP because the measured pressure decay was faster without ECC (6007) than with ECC (6400), which was the opposite of what was expected.
The WRAP calculations predicted a faster depressurization for Test 6007 than for Test 6406, in agreement with the GE test results (Fig. 4).
WRAP satisf1ctorily calculated both the TLTA experiments, within the constraints of the EM specifications.
The Moody critical flow model with a discharge coefficicent CD=
gave higher break flows and therefore more rapid depressurization than measured.
The WRAP calculated peak clad temperatures were above the measured data and showed the same general behavior toward the end of blowdown.
At early times in the transient, WRAP predicted anomalously high clad tempera-tures because of the Barnett and modified Barnett critical heat flux (CHF) correlations specified in the INEL input.
A short run with the Hench-Levy (GE) correlation gave more reasonable results aad this correlation is recommended for future calculations.
IV.
RECOMMENDATION FOR WRAP-BWR-EM USE BY NRR Development and verificaton of the WRAP-BWR-EM system is now complete and it is recommended for use by NRR.
In early CY 80, the WRAP system was made operational on the Harry Diamond Laboratory computer, located in a Maryland suour0 of Washington, and a special remote terminal was installed in Bethesda so that NRR personnel could perform their own calculations with WCAP.
An indoctrination class in the use of WRAP was also held in Bethesca by SRL personnel in early CY 80.
Currently, most of the NRR use of WRAP is being performed at SRL.
These include a BWR-LOCA sensitivity study and other audit calculations, funded by NRR.
RES intends to supply maintenance of the WRAP system over the next few years, to improve modeling and to solve any coding problems that are identified by NRR use.
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'N Robert B. Minogue, Director Office of Nuclear Regulatory Research
Enclosures:
1.
Appendices I and II 2.
Figures 1 through 4 cc:
D. F. Ross, NRR P. Check, NRR T. Speis, NRR G. W. Knighton, NRR R. Audette, NRR
1 REFERENCES 1.
M. R. Buckner, et al., "The BWR LOCA Analysis Capability of the WRAP-EM System," NUREG/CR-0713, Savannah River Laboratory, 1979.
2.
Division of Technical Review, NRC, "WREM:
Water Reactor Evaluation Model," Revision 1, NUREG-75/056, 1975.
3.
Memorandum from E. Case to S. Levine (NRC), "NRR Requirements for LOCA Analysis Computer Programs," RR-NRR-77-5, June 23, 1977.
4 R. R. Beckmeyer, et al., " Users's Guide for the BWR LOCA Analysis Capability of the WRAP-EM System," NUREG/CR-0714, Savannah River Laboratory, 1979.
5.
D. A. Sharp, "The BWR Steady-State Capability of the WRAP-EM System,"
NUREG/CR-0712, Savannah River Laboratory,1979.
6 Memorandum from, L. Shotkin to S. Fahic (NRC), " Response to 11/13/79 Merc, 'In-house Audit Capabilities of Vendors Accident Analyses,'"
December 8, 1979.
7.
Letter from P. Nortn (INEL) to R. Tiller (INEL), " Completion of WRAP-BWR Tasks," October 19, 1979.
8.
C. E. Beyer, C. R. Hann, D. D. Lanning, F. E. Panisko, and L. J. Panchen, "GAPCON-THERMAL-2:
A Com:: uter Program for Calculating the Thermal Behavier of an Oxide Fuel Rod."
BNWL-1897 and BNWL-1898, Battelle Pacific Northwest Laboratories, Richland, Washington, (November 1975).
9.
"RELAP4/M005 - A Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear Reactors and Related Systems - User's Manual," Report ANCR-NU4EG-1335, Idaho National Engineering Laboratory, Aerojet Nuclear Company, Idaho Falls, ID.
(1976).
10.
J. G. M. Anderson, et al., "NORC00L-1, A Model for Analysis of a BWR Under LDCA Conditions, a Revised Report." Report No. NORHAV D-47, Riso National Laboratory (Denmark), (August 1977). Also, NORHAV 0-33 (1977)
.and NORHAV 0-72 (1978).
l 11.
D. R. Evans, "The M0XY Digital Computer Program for Boiling Water Reactor l
Core Thermal Analysis," Report RE-A-77-081, Idaho National Engineering Laboratory, (September 7,1977).
12.
S. R. Fischer, et al., "RELAP4/M006 - A Comouter Program for Transient Thermal-Hydraulic Analysis of Nuclear Reactors and Related Systems,"
Report No. CDAP TR 003, Idaho National Engineering Laboratory, EG&G Idaho, Inc., Idaho Falls, ID.
(January 1978).
[
13.
" General Electric BWR Thermal Analysis Basis (GETAB):
Data Correlation and Design Aoplication," Report No. NE00-10958, (November 1973).
I r
2 14.
G. W. Johnson &nd L. H. Sullivan, "LPWR 'EM' Models for RELAF4/M005,"
Report No. CDAP-TR-78-635, Idaho National Engineering Laboratory, EG&G Idaho, Inc., Idaho Falls, ID (August 1978).
15.
10 CFR, Part 50, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water-Cooled Nuclear Power Reactors," Federal Register, Vol. 39, No. 3 (January 1974).
16.
Memorandum from T. Murley to R. Mattson (NRC), April 4, 1979.
Submittal of WRAP-EM Package Models for 055 Approval," April 14, 1979.
17.
R. Reed. " Verification of the M0XY Code in WRAP for Thermal Analysis of Fuel Assemblies in a BWR During a LOCA," DPST-80-309, Savannah River Lab, 1980.
18.
P. B. Perks, " WRAP Verification with Companion RELAP4/M005 - MOXY Calcula-tions (INEL) and TWRAM-M0XY Calculations (SRL) of a Hope Creek LOCA,
DPST-79-599, Savannah River Laboratory,1979.
19.
F. Beranak, " WRAP-EM BWR/6 Verification Study," OPST-80-484, Savannan River Laboratory 1980.
20.
R. Reed, " WRAP Verification Using TLTA Tests 6007 and 6406," OPST-80-420.
Savannah River Laboratory,1980.
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APPENDIX I:
DESCRIPTION OF CODE INTERFACES A.
GAPCON/ WRAP In the WRAP-EM computational system, GAPCON is used to determine fuel rod conditions at the beginning of the LOCA analysis.
These initial conditions are functions of power level, burnup, fill gas pressure, etc.
The data from the GAPCON calculation are transferred via the GAPCON/RELAP interface to RELAP which calculates the blowdown phase of the LOCA.
The data transfer between GAPCON and RELAP is not straightforward because the fuel models in the two codes differ.
For example, GAPCON mocels a single fuel pin allowing up to 20 axial nodes for detail.
RELAP, on the other hand, models the complete core as one or two (or possibly three) stacks of heat slabs.
One stack may represent the hot bundle; while, the other models the remainder of the core.
Thus, the data transferred from GAPCON to RELAP must be collected, sorted, and interpolated by the interface before being used in RELAP.
Data transferred between GAPCON and RELAP is specified for a given pin (s).
For example, hot channel and average channel heat slab conditions may be determined by " typical" hot rod and " typical" average fuel rod calculations done with GAPCON In the interface, the GAPCON " typical" pin data may be scaled by the number of fuel rods per stack to obtain RELAP heat slab data.
The data transferred by the interface includes the fuel rod geometry (before and after burnup), power, fuel density, ar.d gap heat transfer f actors including fission gas composition and gr:-moles of fission gas.
Where GAPCON data differs as a function of axial location, the RELAP heat slab data are linearly interpolated from the GAPCON data.
8.
GAPCON/MO.VY Initial fuel conditions as calculated by GAPCON are passed automatically to M0XY by the GAPCON/MOXY interface.
As with the GAPCON/RELAP inter-face, the data transfer is not straightforward because of the model differences between GAPCON and MOXY.
GAPCON models in detail a single fuel pin as a function of power level, burnup, fill gas pressure, etc.
GAPCON allows up to 20 axial nodes.
On the other hand, a MOXY calculation is based on a horizontal plane in a fuel assembly where the assembly contains a 7x7 or 8x8 square fuel pin array.
Diagonal symmetry is assumed in the fuel array, but each pin in the remaining half assembly may differ in power and thermal character-istics.
A GAPCON calculation for each pin of the MOXY model is required to properly initialize the MOXY pin array.
In the case of a 7x7 fuel assembly, this requires 28 GAPCON calculations, and for the 8x8 assembly, 36 calculations for a given operating condition.
Since MOXY models a planar section of the assemtly, the data transferred from GAPCON to MOXY are data for the hottest axial level.
The data transfer I-l
includes a check to ensure that the hattest axial node for each GAPCON calculation is at the same elevation.
The data transferred via the GAPCON/M0XY interface includes the fuel rod geometry (before and after burnup), power, fuel density, and gap heat transfer factors including fission gas composition, gm-moles of fission gas, and temperature jump distance.
C.
RELAP/M0XY Transient hot assembly osta during blowndown are transferred from RELAP to MOXY via the RELAP/MOXY interface.
RELAP describes the transient response of the reactor through blowdown during a postulated LOCA.
RELAP typically models the complete core z, one or two stacks of heat slabs.
One stack may represent the hot assembly; the other the remainder cf the core.
MOXY, on the other hand, is used to analyze a planar section of a fuel assembly during a LOCA.
Normally, MOXY, is used to analyze the hot plane in the hot assembly.
Thus, data for the rottest core heat slao are passed to M0XY.
The RELAP/MOXY interface provides M0XY with input of time-decenaent data fcr the ncrmalized power, for the cladding-to-coolant heat trans'er coefficient. and for the fluid temperature durinc the bicadown p;rtion cf the transient.
Other quantities passec to MCXY incluce tne times at which the following events occur in the blowdown calculation:
End of lower plenum flashing which is defined as the time at which o
dryout of the hot bundle occurs, o
Core sprays at rated flow, and
'o End of critical flow at the break.
The above times are used in applying the 10 CFR 50, Appendix K (Ref. 1E) rule for specifying convective heat transfer coefficients for BWR fuel under spray cooling.
D.
RELAP/NORC00L Most of the input data required for NORC00L is available directly (or may be calculated) from the data base created by execution of RELAP.
The correspondence between RELAP and NORC00L variables is provided in Table III of Reference 1.
The interface module INTNOR carries out the transfer of data from the RELAP data base into templated input records for NORC00L. Various algorithms are built into module INTNOR.
A renodalization algorithm allows the user to subdivide a RELAP volume into several identical NORC00L nodes merely by identifying the RELAP volume name for a series of NORC00L nodes.
Each of the nodes is assigned the same state variables characteristic of the original volume.
Since the nodalization of the I-2
1 diffuser region in NORC00L may not be dimensionally consistent with the RELAP nodalization, several correction operations are built into the code:
an equal-mesh transition zone may be created between the upper downcomer and the top of the diffuser; the top of the steam separators is equalized with the bottom of the steam dome.
Nodes in the core and bypass regions are paralleled, as are those in the lower plenum and the region below the diffuser.
In a typical LOCA calculation, NORC00L is executed when end of critical flow at the break occurs in the blowdown.
At that time, the void fractions are essentially unity everywhere in the system except in the nodes close to the spray injection point and in the lower plenum.
To model tnis situation, the interf ace defines a pseudo-water level at the bottom of the lower plenum, by volume averaging and collapsing void fractions required for NORC00L to function properly.
The void fraction is defined as zero below the pseudo level and unity above it.
In all other nodes, the true void fraction is extracted from the data base.
E.
NORC00L/M0XY Ine data interface between NORC00L and MOXY consists of the following quantities:
the hottest plane number, its elevation from the core cottom, and the time at wnica the quench front in the core reached the hottest plane.
The latter time is used in M0XY to complete the specifi-cation of convective heat transfer coefficients according to Appendix K rule 0
I-3
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APPENDIX II:
COMPARISON OF GE AAG V4AP CLA0 TF*PERATUDE FOR BWR/S LOCA The difference in peak-clad tamperature between 99AP and GE is shown in Figure 3.
The obvious deviation at 10 seconds indicates that GE calculates departure fr6m nucleste boiling (DE8) at the highest pow 9 red plane earlier than VRAP.
In addition, lower plenum flashing does not seem to rewet the plane, indicating The the presence of a beat transfer lockout in their feel analysis program.
WRiP (MOXY) analysis shows that the highest powered piene remains ir, nucleate boiling until well af ter lower plenum flashing.
DNS is not reached until 48 seconds at which time tre hot plane uncovers and remains uncovered until quenching at 216 seconds predicted by NORC00L.
The difference in temperatures at early times (+10 secor.ds) is thought to be dun to cnre flow discrepancies.
WRAP craputes a core flow spike at about 9 seconds vnile the GE analysis shows flow stagnation and reversal.
This flow differ 9nce vill r.ause a considerable variation in the heat transfer and it ocents. inst b& fore the GE analysis indicates DN9.
The reverse flow seen by CE is n rticularly 1,nnrtant since it is not clear hnw it affects the critical heat flux cnrrelation they use.
GE employs the General Electrical Critical Quility (V,) - P.niling Length Correlation 2
(GEXt) which is apparently invalid at low mass flei cates (<1 x 10" lbm/hr-fc )
and reverse flow.
Thus, their calculated CHF furiro these ficw repimes may be very low.
The WPAP calculation used the Hench-Levy C:iF correlation since the input date required for the GEXL correlation were net availzble at that time.
A sub-sequent WRAP calculation amoloying tne GEXI. correlation (using 3.0-inch heat slabs as recommended by NRC) indicated that the neat flux at the highest powered slab remained wall under the CHF through the 10 reactnr seconds that the problem was executed.
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