ML17328A380
| ML17328A380 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/13/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17328A379 | List: |
| References | |
| NUDOCS 9008170051 | |
| Download: ML17328A380 (10) | |
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UNITED STATES NUCLEAR REGULATORY COMMlSSION WASHINGTON, O. C. 20555 ENCLOSURE 1
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO THERMAL-HYDRAULICANALYSIS USING THE COBRA IIIC/MIT2 COMPUTER CODE INDIANA MICHIGAN POMER DONALD C.
COOK NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-315 AND 50-316
1.0 INTRODUCTION
Safet Evaluation By letter dated January 30, 1989 Indiana Michigan Power requested NRC review of the report entitled "AEP Reactor Core Thermal-Hydraulic Analysis Using the COBRA IIIC/MIT-2 Computer Code".
The American Electric Power (AEP) Service Corporation has developed the thermal-hydraulic analysis capabilities using COBRA IIIC/MIT-2computer code to analyze the core conditions of the Donald C.
Cook Plant.
The methodology will be used for plant operational
- support, licensee event report evaluations, and other FSAR type analyses as well as for reload evaluations and license submittals.
The model has been used to determine the hot channel fluid conditions and the resulting minimum departure from nucleate boiling ratio (MDNBR).
The accuracy of the model has been verified using three steady state and three transient DNB analyses from Cycle 1 of the Cook Nuclear Plant Unit 1.
The study showed that the results obtained using AEP methods are in excellent agreement with those presented in the licensing documents.
- 2. 0 EVALUATION The core thermal-hydraulic analysis is performed to calculate the coolant conditions in order to verify that the fuel assemblies of the core can safely 9008i7005i 90QSl3 PDR ADOCK 050003i5 P
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2 meet the limitations imposed by departure from nucleate boiling (DHB).
The thermal-hydraulic methodology described in the topical report is based on a
single stage method.
A similar methodology has been approved for use by the Virginia Electric Power Company(VEPCO}.
The COBRA-IIIC/MIT 2 computer code calculates the flow and enthalpy within interconnected flow channels by solving finite difference equations of continuity, energy, and momentum.
The mathematical model is applicable to both steady state and transient conditions and the model considers both turbulent mixing and diversion crossflow.
In formulating the model one-dimensional, two-phase, separated, slip-flow is assumed to exist during boiling.
The two-phase flow structure is assumed to be fine enough to specify the void fraction as a function of enthalpy, flowrate, heat flux, pressure,
- position, and time.
Sonic velocity propagation effects are not included.
Mithin a
- channel, the diversion crossflow velocity is assumed to be small compared to the axial velocity, to allow the use of a simplified equation for the
'conservation of transverse momentum.
The same finite difference equations are used for both steady state and transient computations.
Initial conditions are obtained by performing steady state calculations and then the transient calculation is performed.
Time dependent forcing functions consisting of inlet temperature, inlet flow, system
- pressure, and core average heat flux are used to establish boundary conditions at succeeding times.
The calculation iterates over the first time step until the flow solution converges.
The converged solution is then used as the initial condition for the new time, and the procedure continues for all the subsequent time steps.
Although the equations of continuity, energy, and momentum form the basic structure of the mathematical model, their solution is still dependent upon the use of empirical correlations.
The COBRA IIIC/MIT2 computer code allows user specification of the appropriate correlations.
2.2 Models and Correlations The void fraction of the flow was calculated using Levy's subcooled void fraction model.
The single and two-phase pressure drops are calculated by using the isothermal friction factor correlation.
The friction factor is corrected for wall viscosity.
For methodology verification, the M-3 correlation in conjuction with F-factor correction was used.
The grid and unheated wall correction factors are applied to predict the critical heat flux.
The number of axial intervals used for both the steady state and the transient analyses was 42.
This is considerably fewer than the 156 for steady state and 78 intervals used for transient analyses by Virginia Power in their verification of the methodology.
In a letter dated June 22, 1990 Indiana Michigan Power provided the results of a sensivity study which AEP performed to show that the number of axial intervals used (42) is sufficient.
The study involved two steady state and two transient cases.
For each of the four cases the minimum DNBR was calculated using 42, 60 and 156 axial intervals.
The results were within 2.6X of the original calculation for all cases.
Thus it was concluded that 42 axial intervals is sufficient.
2.3 H draulic Model Descri tion Eighth core symmetry is assumed, and thus a 1/8 core segment is modeled.
The location of the hot assembly is assumed at the center of the core.
The hot assembly is modeled as an array of subchannels, while the remaining assemblies are modeled as an array of lumped channels.
For the Unit 1 core analysis, a 58 channel model was developed.
It consisted of 30 lumped assembly channels and 28 subchannels.
Each of the lumped assemblies were modeled using a lumped flow area, lumped heated and wetted
perimeters and an effective gap.
The central assembly was modeled in greater detail, consisting of four different types of subchannels perimeter cell, corner cell, thimble cell and unit cell.
2.4 Thermal Model Descri tion The thermal model consists of an inlet flow distribution, radial and axial power distributions, and appropiate reactor operating conditions.
For analysing the transients, the time dependent forcing functions of system
- pressure, inlet flow, inlet temperature, and core power are also specified.
Thermal hydraulic design parameters form the basis of the model.
The radial power distribution is based upon the design value of F~
and the axial power 6H'istribution is based on the reference axial shape.
The thermal design flow rate is used in determining the core average mass velocity, and the thermal hydraulic design values for inlet temperature, system pressure and power level are used as operating conditions.
The thermal model is then imposed upon the hydraulic model in order to obtain the complete thermal-hydrauliuc representation of the core.
Since this representation is dependent upon thermal-hydraulic design parameters, revised representations must be considered in the event of any subsequent design changes.
In general, the hydraulic model remains relatively fixed since it is affected only by changes in the mechanical design of the fuel. However, the thermal model can be significantly affected by changing any one of the design parameters.
The inlet flow distribution used in the 58 channel model assumes a
5 percent flow reduction to the hot assembly while the peripheral assemblies have a flow fraction slightly greater than l.0.
The average of all the flow fractions is approximately 1.0.
In the radial power distribution for Cycle 1 the hot assembly as well as the adjacent assemblies are given relative powers of 1.475, while lower relative powers are assigned to the remaining assemblies.
The average of all the assembly relative powers is 1.0.
The reactor operating
conditions include core power level, core flow rate, core inlet temperature, and operating pressure.
The operating conditions for the analysis are obtained by applying the maximum steady state errors to the rated values.
For performing the transient analysis, the COBRA IIIC/MIT-2 computer code requires the four normalized forcing functions, namely pressure versus time, enthalpy or inlet temperature versus time, average mass flow versus time, and channel power versus time.
These forcing functions were obtained from the Unit 1
FSAR for the comparison study.
They can also be obtained from transient analysis performed using a thermal hydraulic systems
- code, such as RELAP5.
Indiana Michigan Power will obtain the forcing funmctions from another source since they do not have approval for use of RELAP5 at present.
2.5 En ineerin Uncertainties After formulation of the overall thermal-hydraulic representation of the core, engineering uncertainties are then applied to account for manufacturing tolerances used in the fabrication of the fuel.
These fabrication tolerances are assumed to occur in the hot channel, and are called hot channel factors.
E The heat flux engineering subfactor F~ is a ratio of local maximum at a point to core average heat flux.
The Focused to evaluate the maximum heat flux, is determined by statistically combining the tolerances for the fuel diameter,
- density, enrichment and fuel rod diameter, pitch and bowing.
The value of 1.03 is used.
The enthalpy rise engineering subfactor is defined as a ratio of maximum to core average enthalpy along a channel.
The value of 1.02 is used.
2.6 Thermal-H draulic Model Yerification 4
Three steady state and three transient analyses were performed to verify the calculational accuracy of the AEP methods.
The COBRA IIIC/MIT-2code input was developed for each of these cases by using the reactor core data provided in
original analysis(documents.
The steady state analysis and comparison of the results is shown below.
Surry Core Analysis at 100% Power VEPCO 1.94
'AEP 1.90 3-Loop PWR Analysis at 118% Power ANF 1.985 AEP l.86 Cook Unit 1 FSAR at 100K Power FSAR 1.97 AEP
- 1. 93 The results of the transient analyses are as follows.
Excessive Load Increase Transient FSAR 1.56 AEP 1.57 Uncontrolled Control Rod Assembly Mithdrawal at Power Transient FSAR 1.40 AEP 1.42 Complete Loss of Reactor Coolant Flow Transient FSAR
- 1. 38 AEP 1.40 As can be seen by the comparison of data, the agreement is excellent.
In all cases the AEP results are either conservative or insignificantly different from the previous calculation of minimum DNBR.
- 3. 0 CONCLUSION Based on our review of the methodology presented and the results of the comparison studies, we conclude that the use of the AEP reactor core
~ I
thermal-hydraulic analysis using the COBRA IIIC/MIT-2 computer code is acceptable.
This methodology,may be used for the Donald C. Cook Nuclear Plant Units I and 2 only.
Principal Contributor:
M. Chatterton
. Date:
August i3, 1990
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