Post on 15-Apr-2018
INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU Reactor
IAEA Workshop on Advanced Code Suite for Design, Safety
Analysis and Operation of Heavy Water Reactors
2012 Oct 2 - 5
Institute for Nuclear Research
Pitesti, Romania
I. Patrulescu
INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
2
OUTLINE
1. Historic background of physics, thermal hydraulic
and fuel performance codes used in INR
2. Three Dimensional Diffusion Code DIREN
3. Programming features of DIREN
4. Code System Verification and Validation
5. WIMS-DRAGON-DIREN-RELAP Sample Results
6. Fuel performance codes presently used in INR
INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
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1. Historic background of physics, thermal hydraulic
and fuel performance codes used in INR
AECL reactor physics, thermal hydraulics and fuel performance
codes were available to INR after 1978 and were used on a small size
main frame: CDC CYBER 170/135.
These reactor physics codes were:
-PPV (cell code), MULTICELL (super cell) RFSP, CHEBY,
CERBERUS, FMPD, CEBXEMAX (reactor core).
The AECL thermal hydraulics codes available were FIREBIRD III,
HYDNA-2, HYDNA-3, NUCCP (NUCIRC).
The fuel performance code was ELESIM.
In parallel INR had access to other computer codes from
international libraries (e.g.):
-WIMS (cell), CITATION (reactor core), TWOTRAN (2D transport
calculation) , MCNP (Monte Carlo approximation to transport equation).
INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
When PC’s appeared, some of the above codes were ported from
mainframes to the new personal computers, e.g.:
-PPV, MULTICELL, FIREBIRD III, HYDNA-2, HYDNA-3, NUCCP (NUCIRC).
In the same time work was done to create INR own computer codes.
Reactor physics codes written were: CP2D for the cell , PIJXYZ for the
reactivity devices incremental cross sections and DIREN for reactor core (3D
multi group diffusion). PIJXYZ solve the integral transport equations using first
collision probabilities. CP2D and PIJXYZ were developed by M.Constantin
et.al.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Some AECL codes in executable form were available to use in INR, in
the framework of INR – AECL agreement for scientific cooperation and in
certain condition. These were:
-WIMS-AECL-IST for CANDU cell calculations,
-RFSP-IST for reactor core,
-CATHENA-IST for thermal hydraulics,
-ELESTRES-IST for fuel performance evaluation.
-ELOCA-IST
The codes used in fuel performance, including INR created or general
codes adapted for CANDU, are described in section “Fuel performance
codes presently used in INR”.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
2. Three Dimensional Diffusion Code DIREN
AECL reactor physics design and core performance evaluations
were done for CANDU reactors using the following basic steps:
i)-”Cell calculation”. Homogenized macroscopic cross sections
are generated for the basic cell (fuel bundle and corresponding
moderator).
ii)-”Reactivity device homogenization”. Incremental macroscopic
cross sections due to perturbations induced by reactivity devices
iii)-”Reactor core calculation”. Using the above calculated cross
sections a three dimensional finite difference approximation to diffusion
equation code.(RFSP, Frescura and Wight, 1982) was used to evaluate
Keff and flux distribution
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
For evaluating CANDU- Cernavoda reactor in INR, initially
(1978) and for some time (1990) the AECL reactor physics
calculations were done using POWDERPUFS,-MULTICELL-RFSP
codes system. In the same time alternately WIMS cell code was
used for cell calculations and PIJXYZ for reactivity devices
incremental cross sections.
The same basic steps were used in INR for reactor physics
calculations and in later code development. This successive steps of
homogenization and energy group collapsing is the basis of standard
approximation in reactor physics.
For RFSP, FMDP (for which the author was responsible in INR)
instead of porting to the new FORTRAN compiler on PC’s the
creation of new diffusion code from scratch was preferred.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
This was done for the following reasons:
-relative difficulty of porting process due to peculiarities of CDC
CYBER, e.g. 60 bits words;
-easier and more flexible use of a own created code,
-simplicity of basic equations to be solved,
-inexpensive verification and validation by comparison to other
classic diffusion codes (in our case CITATION was used),
-possibility of easily implementing new approximations and
methods.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
The work at DIREN began after the PC were introduced in use
(by 1992). DIREN intended to be an inexpensive solution to reactor
core calculation. Beside the author a large and essential contribution
to the writing of this code was brought by Mr. .
DIREN is a multi group, finite difference approximation of
diffusion equations. The general form of diffusion equation was used.
For solving the time dependent diffusion equation the quasi static
approximation was applied. DIREN has the capability of solving the
diffusion equation in 2 or more (up to 10) energy groups taking into
account the up scattering. The approximations and methods used are
classical in reactor physics computation: finite difference
approximation of diffusion equation and related iterative solutions of
eigenvalue problem.
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V.Raica
INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Due to CANDU reactor core complexity (a lot of reactivity
devices perpendicular to horizontal fuel channels) three layers of
arrays are used:
-fuel channel,
-control devices array,
-mesh points positions.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
A typical diffusion calculation has the following five basic steps:
-1) Cross sections generation for each fuel bundle in the core.
The procedure depends on the type of calculation. For example,
in the case of a simulation, cross sections are calculated by
interpolation as function of irradiation in tables that are given in input.
-2) Obtaining cross section at each mesh point.
This is done firstly by assigning reflector cross sections at all
mesh points. Next fuel cross sections generated at step 1) are
allocated at mesh points that overlap fuel bundles array.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Volumetric weighing is used when mesh point position differs
from fuel array dimension. Finally incremental cross sections are
added to the volumes affected by each device.
-3) Finite difference discretization.
Diffusion equations are transformed in linear numerical equation
by integrating over mesh point volume and approximation leakage
terms by finite differences. Coefficients of finite difference form of
diffusion equation are calculated at each mesh point using cross
section allocated at each mesh at step 2.
-4) Iterative solution of eigenvalue problem.
The set of linear numerical equations constitutes an eigenvalue
problem.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
This problem is solved by using a source iteration procedure; in
the case of a multi group with up-scattering inner iterations are used.
After the convergence is attained basic results (Keff and relative group
flux distributions) are calculated.
-5) Obtaining the final primary results.
Group fluxes for each fuel bundle in the core are calculated from
mesh points flux distributions. Fuel bundle fluxes are normalized to
actual core fission power using production cross sections and power
to flux ratios for each bundle. In the process the power distribution for
every fuel bundle is calculated. There are also specific data, e.g.
reactivity core decay rate with burn-up, increase in core reactivity at a
fuel bundle replacement, that may be evaluated using the basic
group flux, power and cross sections distribution on the fuel bundles
array.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
There are some types of calculations that are usually performed
for CANDU reactors incorporating the above described basic steps:
Direct simulation on discrete time steps.
Beside the direct time step simulation another usual
approximation for CANDU core is the AECL proposed “time-average”
calculation. This approximation is used to evaluate a refueling
equilibrium situation in CANDU.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
From the point of view of the thermal-hydraulics parameters that
have a significant effect on cross section (fuel temperature, coolant
density, fuel bundle power) the following calculating methods were
developed:
Straight approximation;
Core average values are used for the thermal-hydraulics
parameters mentioned above.
Local parameters approximation;
For each fuel bundle above mentioned parameters are used in
cross sections evaluation. These may be obtained in an interactive
process with a specialized thermal-hydraulic code or reading them
from given files.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Local parameters with history for each bundle;
In this case cell code is executed for each bundle with thermal-
hydraulic parameters as specific input data. Besides irradiation other
data are to be kept in memory for each bundle in order the cell code
to run correctly. For the computer memory available now this is no
longer a problem WIMS can be used for this type of calculations.
Local parameters option is the straightforward, “correct” approach.
Coupling with cell and super cell codes was first done with WIMS-
PIJXYZ and later with WIMS-DRAGON.
As the fuel bundle power is a parameter that affects the cross
sections and a product of calculations an iterative procedure with
steps 1 to 4 is used until the axial power distribution is converged. In
case of cell code calculation at each bundle this lead to relative large
calculations time.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Other capabilities of DIREN are:
-Calculation of higher modes of diffusion equation. Up to 14
modes that are solutions to the same diffusion equation can be
obtained.
-Flux mapping given the flux detector readings and modes
generated previously. These modes are the higher modes that are
solution to the diffusion equation and flux distribution for typical
configurations that can occur in operation.
-Simulation of xenon effects. Usually the xenon effects are
accounted for using equilibrium values. This option can treat xenon
effect by taking into account xenon concentration dependence of
local powers for each bundle.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
-Global and local simulation of liquid zone control. This
calculation module allows to search for all zone liquid controllers
positions which give a required reactivity in global option or the
individual position of positions which give the closest detector
readings to the specified set of readings.
-3D spatial kinetics solution in quasi static approximation;
-Core burn-up simulation with automatic refueling.
-Coupled physics and thermal hydraulic calculations.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Coupling with other codes was first done with WIMSD5-B3 and
PIJXYZ. WIMSD5-B3 cell code was available in source form from
NEA library and super cell code developed in INR.
Later DIREN was coupled with WIMS-DRAGON-RELAP. The
DRAGON 3.05E, developed at Montreal Ecole Politechnique by G
Marleau et.al., available in source form in “open source” license, was
used because it allowed obtaining the incremental cross section
directly in one run.
This code can also be applied to other types of thermal reactors.
Actually the first coupled reactor physics and thermal hydraulic with
RELAP5-MOD3 was done for a PWR reactor namely SPERT-III for
which a lot of transient measurements are available. RELAP is
available only as executable, it is developed at Idaho National
Engineering Laboratory, USA.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
DIREN can be used in standard command line or with a
graphical user interface. Beside execution and printing options that
can be introduced before the execution is started the user can
examine the results as the execution of the code proceeds. Types of
graphics that can be displayed are:
-volume over which the incremental cross sections are added for
a selected device,
-one dimensional flux distribution over a selected direction and
plane the user select,
-two dimensional flux distribution in section xy, xz, yz for a plane
selected by user,
-colors associated with flux values in a selected cut (color map).
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Auxiliary programs were developed which can be used to:
-generate kinetics parameters from nuclear library data,
-generate xenon parameters based on nuclear data library,
-generate relevant fuel history irradiation results starting from
primary data provided by DIREN
-couple cell codes WIMS and DRAGON to DIREN,
-couple thermal hydraulic code RELAP to DIREN.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
3.Programming features of DIREN
The programming of DIREN and auxiliary programs were done
generally under versions of FORTRAN that were developed for PC’s.
Graphical interfaces were developed under C++ with mixed
FORTRAN/C++ projects. VISUAL BASIC was only used for
irradiation history auxiliary code.
The programming was done based on the following intents:
-Apply programming principles that make source code easier to
follow and debug;
-Keep the data in the memory for the major calculation
processes, especially in the iteration;
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
-Build a simple input interface and easier to use for introducing
data (e.g. each type of data have a preceding comment line
describing the type and format);
-Introduce the graphical interface but only for execution and
printing options. The source code was the same as in command line
and this graphic interface can be used optionally;
-Dimensions of variable in “COMMON” could be easily changed
by dimensions only in one file; that file is introduced by INCLUDE in
any routine that has COMMON blocks;
-Coupling was done in a strait forward (but not necessary the
best) way. The interfacing between DIREN and other codes (cell and
thermal hydraulic) was done using text files and small buffer
programs.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
A buffer small routine is dedicated to a given code (e.g. for
coupling cell code WIMS is used a different routine that for
DRAGON) .
This simple approach to coupling was imposed partly because
some codes are available only in executable form or the changes in
source code were intended to be kept at a minimum.
Thus in DRAGON no intervention in the source was done and
code was used as recommended by authors in the Cygwin interface
to WINDOWS. An attempt was done to port DRAGON to
FORTRAN PC but this gave slightly different results, probably due to
the way some assembler routines that manage memory allocation
were emulated.
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In WIMSD-B3 only minor modifications were done to write cross
sections and calculate H factors (power over flux ratios) used in flux
normalization to reactor power.
The coupling to RELAP was done using an dedicate buffer
program (BPR) which is run from within DIREN. The data transferred
to BPR are:
-power distribution for all bundles in the core,
-thermal hydraulic map in the core (fuel channels assignment to
each RIH to ROH pass).
Buffer program uses interface text files to transmit data between
DIREN and RELAP.
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Starting from a specified file and data taken from DIREN, BPR
creates input data file. Then launches RELAP execution and
retrieves data that is expected by DIREN: average fuel temperature,
average coolant temperature and average coolant density.
The data is retrieved from RELAP restart file using STRIP input
card which writes selected data in a file named stripf. Consistency
tests showed that, besides the data built in restart file, RELAP
requires also the temperature distribution is source. Using the above
procedure that values were extracted and written in input data in
source cards for each restart step.
The use of text files to interface DIREN to RELAP does not
create problems because the amount of data transmitted is not very
large.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
For the single threaded compilers no special problems occurred
when calling cell code and writing interface text files repeatedly,
especially in cell code call for each bundle. With appearance of multi
threading FORTRAN compilers some problems occurred due to
delayed writing of files. Solution found was to check for files
availability in certain conditions.
The number of mesh points used in DIREN core model is
56x56x52. For the usual PC speed of 2.7 Ghz this leads to a 0.05
sec/iteration in two group calculations and at 0.2 sec/iteration for 7
group. A WIMS cell calculation requires around 0.5 seconds. These
values make local parameters calculation with cell calculation for
each bundle possible.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
4. Code System Verification and Validation
First step in verification was to compare results for basic solution
to the diffusion three dimensional multi group equations using simple
test cases with classic diffusion code CITATION. A comparison was
also made with results of FMDP.
For complex comparison a CANDU specific test problems
proposed in the framework of IAEA contract ( 1996, 1990) were used.
Configuration is a simplified 600 MW PHWR CANDU reactor with
380 horizontal fuel channels surrounded by moderator. A channel
has 12 axial fuel bundles. A limited number of control devices are
between fuel channel and perpendicular to them. The following
neutron absorption devices, typical for this CANDU reactor, are
represented in model: adjuster rods, liquid zone controllers and shut-
off rods.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
The cell codes WIMSD5-B3, DRAGON3.05b and libraries used
in INR were compared with benchmark problems gathered in IAEA
WLUP project (WIMP Library Update).
Time dependent solution was verified by comparison with test
problems run with CERBERUS. Also results obtained with DRAGON-
DIREN-RELAP were compared with experiments that were
performed to investigate reactivity accidents on SPERT-III (Special
Power Excursion Test).
Results obtained with WIMS-DRAGON-DIREN for a 130 days
operation history of CERNAVODA unit 1 were compared with similar
results with RFSP that Cernavoda made available to us.
DIREN coupling with RELAP was verified by concistency tests
and by comparison with previous results obtained in INR.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
5.WIMS-DRAGON-DIREN-RELAP Sample Results
The reactor physics code system WIMS-DRAGON-DIREN-
RELAP was used in INR to obtain result similar to those in Design
Manual Reactor Physics.
It was also applied to retrieve relevant data for irradiation history
which are used by fuel performance analysis:
- burn up and power history for selected bundles in the core,
- power and burn up histograms for all bundles in the core at
selected time steps,
-diagrams which shows the number of bundles that are under or
above defect curves.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
WDD system was also applied to evaluate the impact on the reactor
core of using other types of advanced fuels in CANDU: SEU (up to
1.1 %), RU, MOX and Thorium fuels.
In order to couple WIMS-DRAGON –DIREN to RELAP following
thermal hydraulic models were developed:
-Simple boundary condition from RIH to ROH for steady state
calculations. The components between RIH and ROH (inlet feeder,
inlet end fitting …) were given for each channel using design data.
-2 loops with two passes with all significant components in the
loop. Also in the model are pressurizer, ECCS, Feed and Bleed. The
secondary loop in steam generators are simulated only with
boundary conditions at inlet and outlet from SG. For some types of
accidents the model for secondary loop in SG should be extended.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
-Two loops with two passes but with two flow channels for
each pass.
-Two loops with two passes with 9 flow channels for each
pass. The limitation to 9 passes is imposed by RELAP limits.
Anyway the calculation time increases significantly (by a factor of
approximately 4).
The RELAP models are used both with point kinetics and
coupled with WIMS-DRAGON-DIREN.
In case of space kinetics with DIREN two approximations can
be used: interpolated cross sections and cell calculation at each
bundle.
INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Tests done on a RIH break both with point kinetics and
3D space kinetics showed good results.
In the near future it is intended to cover, if possible, all the
transients regimes that appear in accident analysis both with
point and space kinetics and compare the results with
previous results obtained previously in INR.
Following slides are samples of graphic DIREN output the
can be obtained during and/or after a run using graphic
interface.
INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Sample one dimensional flux distribution produced by in line graphic routine in DIREN
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Sample two dimensional flux distribution produced by in line graphic routine in DIREN at plane 31, xy
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Sample color map produced by DIREN
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Evolution of thermal flux shape for a 35 % RIH break
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T=0. sec, LOCA starts T=0.1sec T=0.2 sec T=0.2914sec, ROP activated
T=0.73 sec T=0.8214 sec , SOR in T=1.1128 sec T=1.5492 sec
INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
6. Fuel performance codes presently used in INR
A. Normal Reactor Operating Conditions:
The codes simulate thermal and mechanical behavior of a single
fuel element under normal reactor operating conditions.
ROFEM 1B – use a two-dimensional (radial-axial) finite element
approach (starting from original FEMAXI adapted for CANDU);
ELESIM – utilizes a one-dimensional (radial) finite difference
representation;
FEMAXI6 – uses a two-dimensional (radial-axial) finite element
approach (Japanese code adapted to CANDU;
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
ELESTRES-IST 1.0* – contains one-dimensional models of heat
generation rate, temperature distribution, fission-gas release, and
pellet-to-sheath heat transfer. A two-dimensional axi-symmetric
stress-strain analysis is used to compute the stresses and strains
in the pellet and in the sheath.
* Codes that are available in INR-AECL agreement, presently in
renewal process
B. Reactor Accident Conditions:
CAREB – uses to study fuel performance under postulated reactor
accident conditions (LOCA, RIA);
ELOCA* - uses to study fuel performance under postulated reactor
accident conditions (LOCA, RIA);
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Structural Analyses
ANSYS – Finite element method. Structural analyze of fuel bundle
in normal operating condition.
BUNDLEG – Fuel bundle geometry optimization (developed in
INR)
ROFEM 1B FUEL PERFORMANCE CODE
Main modifications (starting from original basis FEMAXI III):
Microstructure dependent fission gas release;
Temperature dependent grain growth and pellet restructuring;
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Temperature, porosity and burnup dependence of thermal conductivity;
Burnup dependence of the radial power profile in the fuel pellet;
Pellet to clad heat transfer via solid-solid, gas and radiative components;
Possibility of analyzing CANDU type fuel geometry and operational
conditions.
Verification/Validation
Fuel irradiation experiments performed in TRIGA reactor –Pitesti
PIE performed in INR – Pitesti hot cells;
FUMEX blind code comparison exercise.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
ROFEM 1B FUEL PERFORMANCE CODE
MAJOR CALCULATIONS
Thermal processes:
Heat transfer between coolant and cladding;
Heat transfer between fuel and cladding;
Heat generation distribution in fuel;
Temperature distribution in cladding;
Fission gas release, inner gas pressure.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Fuel mechanics:
Thermal strain;
Elasticity;
Plasticity;
Creep
Cracking and healing;
Initial relocation;
Densification
Swelling
Hot pressing
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Clad mechanics:
Thermal strain;
Elasticity;
Plasticity;
Creep;
Anisotropy;
Radiation hardening;
Ridging;
Pellet-clad interaction.
Material properties dependent on
temperature, irradiation.
INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
CAREB is a code that calculates the thermal-mechanical behavior of a fuel
elements under transient conditions (LOCA, RIA).
The principal process modeled are:
Transient thermal behavior.
Fuel thermal expansion, cracking and melting.
Stored heat during transient.
Internal gas pressure variation during transient.
Fuel/sheath heat transfer.
Thermal, elastic and plastic sheath deformation (anisotropy).
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
High temperature Zircaloy-4 sheathing creep behavior.
Beryllium assisted crack penetration of the sheath
Clad damage accumulation.
The new version of CAREB code has been compared against
ELOCA code results.
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
48
0
2
4
6
8
10
12
0 10 20 30 40 50 60 70 80 90 100
TIME (s)
PR
ES
SU
RE
(M
Pa
)
37-element:outer element
43-element: inner element
43 element: outer element
0
2
4
6
8
10
12
0 10 20 30 40 50 60 70 80 90 100
TIME(s)
PR
ES
SU
RE
(M
Pa)
37-element: outer element
43-element: inner element
43-element: outer element
20% RIH INTERNAL GAS PRESSURE. (ELOCA
results)
The comparison of
CAREB predictions to
ELOCA predictions
show a good
agreement between
the codes results.
20% RIH INTERNAL GAS PRESSURE. (CAREB
results)
INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
49
Axial (Z-direction) displacements distribution in RU-43
short fuel bundle end plate (ANSYS results). Single
side-stop case.
ANSYS - Structural analyze of fuel
bundle.
BUNDLEG - CANDU FUEL BUNDLES DESIGN
BUNDLEG code is used to determine a
lots of geometrical fuel bundle variants.
At the end of calculation the designer
could be select the optimal variant.
INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
RELEVANT REFERENCES
J. R. Askew, (1968), “A General Description of the Lattice Code WIMS” J.B.N.E.S., 564
H.C. Chow and M.H.M. Roshd (1980), “SHETAN- A Three Dimensional Integral Transport Code for Reactor Analysis”, AECL-6878, September 1980
M.Constantin et al. ,(1996) “Some Approaches for Improving the Performances of 3D Integral Transport Codes”, in Proceedings of Annual Meeting on
Nucl. Technology, Manheim, Mai 23-25, 1996, p.7
M.Constantin et al, (1995), ” Performances Evaluation of 3D integral transport code:PIJXYZ”, (In Romanian) INR,Pitesti
A. R. Dastur,D.B.Buss (1983) “ MULTICELL A 3-D Program for the Simulation of Reactivity Devices in CANDU Reactors”, AECL-7544
G.M.Frescura and A.L. Wight, (1982), “CANDU-PHW Fuel Management “in “Operational Physics of Power Reactors”, Proceedings of the Training Course
held at Trieste, IAEA-SMR-68/II
E.Lewis, W.F.Miller Jr, (1984) “Computational Methods of Neutron Transport”, Chap.1 John Wiley&Sons,
A. A. Pasanen, (1982), “ Fundamentals of CANDU Nuclear Design” in “Operational Physics of Power Reactors”, Proceedings of the Training Course held
at Trieste, IAEA-SMR-68/II
I.Patrulescu et al.,(1993), ”User’s Manual for 3D diffusion code DIREN”, (In Romanian) Internal Report, INR Pitesti, 1993
Rouben B et al., (1988),“Calculation of 3D Flux Distributions in CANDU Reactors Using Lattice Properties Dependent on Several Local Parameters”,
Nucl. Sci. Eng.,98, 139
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INR Coupled Reactor Physics and Thermal Hydraulics Calculation System for CANDU reactor
Simionovici et al (1984), “Developing of a code based on collision probabilities and interface currents” (in Romanian), Institute for Nuclear Research,
Pitesti
X.Warin, (1993), “Methodes deterministes de resolusion de l’equation integrale du transport neutronique” , EDF 94NJ0057
***, (1996), “In-Core Fuel Management Benchmarks for PHWRs”, IAEA-TECDOC-887,June 1996
***, (1990), “Summary Report of First Research Coordination Meeting on In-Core Fuel Management Programs Related to Core of PHWRs”, Buenos Aires,
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