6 - BARC

KUDANKULAM 1000 MWe VVER
6.
R E A C T O R T E C H N O L O G Y: K U D A N K U L A M
1000 MWe
VVER
INTRODUCTION
Considering the growing energy demands and the necessity to increase the energy potential, a second line of light water reactors has
been added to the current indigenous programme of Pressurised Heavy Water Reactors. Two Light Water Reactors of 1000 MWe VVER
units are being installed at Kudankulam in collaboration with the Russian Federation. These reactors in addition to accelerating the
nuclear energy potential would also help in expanding the knowledge pool by broadening the research activities in reactor technology.
This chapter on Kudankulam VVERs, highlights the recent work in the areas of reactor analysis, code development with visual interfaces
for physics computation and pin-by-pin simulation of hexagonal lattice cores.
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KUDANKULAM 1000 MWe VVER
V V E R is an acronym for “Voda Voda Energo Reactor” meaning
a pressurising system connected to the reactor with each loop
water-cooled, water moderated energy reactor. The VVER
containing a horizontal steam generator, a main circulating pump
reactors belong to the family of the Pressurised Water Reactors
and passive part of emergency core cooling system
(PWRs). The KK-VVER has a three-year fuel cycle. This reactor
(accumulators). The loops are connected with the reactor
requires annual refueling of one third of the core i.e.,
pressure vessel assembly by interconnected piping. The reactor
approximately 55 fuel assemblies.
also consists of a reactor protection and regulation system,
engineered safety features, auxiliary system, fuel handling and
The reactor plant consists of four circulating loops and
storage system.
Major Systems of 1000 MWe VVER
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6.1
VVER-1000 MWe REACTOR ANALYSIS
After generating the complete lattice database with EXCEL code
for 11 fuel types, the VVER-1000 Mwe reactor core of KK Project
The VVER-1000 MWe reactor core of Kudankulam (KK) Project
was followed up for 8 fuel cycles. Each hexagonal assembly cell
is a Pressurized Water Reactor (PWR) of Russian design. It is
was divided into 54 triangular meshes. The results like critical
necessary to develop indigenous capability of in-core fuel
soluble boron, radial and axial power distribution, 2-D and 3-D
management of these reactors. This capability is also essential
peaking factors were compared with Russian data. The calculated
for an in-depth review of the PSAR documents submitted by
critical boron with a uniform keff normalization agreed well with
Russian Federation for KK Project. The detailed analysis and
Russian data for all eight fuel cycles. The deviation was slightly
comparison of results with the Russian design project reports
more for first fuel cycle, possibly due to non-equilibrium Sm
giving the physical characteristics under various steady state
load. Power dependent feedback is being implemented
conditions has revealed that it is essential to develop capability
in TRIHEX-FA code to reduce the deviations observed in power
for analysing some of the slow (xenon) and fast transients.
distribution. The modeling of reflector region is also being fine
tuned.
Indigenous lattice burnup code EXCEL and core diffusion
analysis codes TRIHEX-FA and pin-by-pin simulation code HEXPIN
V. Jagannathan, <vjagan@barc.gov.in>
have been developed and are used to analyze the KK core.
Typical Comparison of Radial Power and Burnup Distribution Core Follow-up Simulation for KK Cycle-8
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KUDANKULAM 1000 MWe VVER
6.2
DEVELOPMENT OF CODE SYSTEM FOR
THERMAL REACTOR DESIGN WITH
VISUAL AID SOFTWARE PACKAGES
‘VISWAM’-A
COMPUTER
CODE
PACKAGE FOR THERMAL REACTOR
PHYSICS COMPUTATIONS
The nuclear cross section data and reactor physics
design methods developed over the past three
decades have attained a high degree of reliability
for thermal power reactor design and analysis.
This is borne out from the analysis of physics
commissioning experiments and several reactoryears
of
operational
experience
of
two
types of Indian thermal power reactors, viz. BWR
and PHWR. Our computational tools were also
developed and tested against a large number of
IAEA CRP benchmarks on in-core fuel management code
package validation for the modern BWR, PWR,
VVER and PHWR. Though the computational
algorithms are well tested, their mode of use has remained
rather obsolete since the codes were developed
when the modern high-speed large memory
computers were not available. The use of Fortran language
limits their potential use for varied applications.
A specific Visual Interface Software as the Work Aid
support for effective Man-Machine interface
(VISWAM) is being developed. The VISWAM package when
fully developed and tested will enable handling of the
input description of complex fuel assembly and the
reactor core geometry with immaculate ease. Selective
display of the three dimensional distribution of
multi-group fluxes, power distribution and hot
spots will provide a good insight into the
analysis and also enable intercomparison of different
nuclear datasets and methods. Since the new
package will be user-friendly, training of requisite human
resource for the expanding Indian nuclear power
programme will be rendered easier and the gap between
an expert and any new entrant will be greatly reduced.
VISWAM Code Package for Reactor Physics &
Shielding Computations
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Typical Screen View of ‘XnWlup’ Software to
View Microscopic Cross sections
Typical Screen View for Input for VVER
Assembly Cell Description
VISWAM: Present Status of Code development
The visual mode of creating input to typical VVER fuel assembly
description is also given. The visual input can be internally
Typical viewing of multigroup cross section by the ‘XnWlup’
transferred to the other digital form of input.
code is shown in figure. One of the combo boxes for creating
EXCEL input file is also shown. There are separate boxes for
entering/editing the input set for pincell, supercell, assembly
diffusion module etc.
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Typical Screen View of Input for VVER
Assembly Geometry
Typical Thermal Flux Profile in a Thorium
Loaded Reactor – by Plot3D
The 3-D flux profile of the complex core simulation can be viewed
for any axial plane and each energy group. Figures give typical
3D flux display at selected plane for different types of reactor
analysis. These were plotted using the program ‘RealPlot3D’ or
the ‘Display’ program. These 3D plots give the clear depiction of
flux profiles in large 3D cores. Inadvertent input error, if any,
can easily be identified and corrected.
Typical Thermal Flux Profile in a LWR
Typical Epithermal Flux Profile in
TAPS BWR – by Plot3D
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V. Jagannathan,
Reactor Technology & Engineering
<vjagan@barc.gov.in>
KUDANKULAM 1000 MWe VVER
6.3
HEXPIN CODE FOR PIN BY PIN SIMULATION
OF HEXAGONAL LATTICE CORES
The code ‘HEXPIN’ has been developed for core follow-up analysis
for first time with a pin-by-pin cell description of the entire core
and reflector regions up to pressure vessel. The input to HEXPIN
code consists only of fuel assembly type disposition.
The geometrical specifications within each fuel type are directly
derived from the output of hexagonal lattice cell burnup code
EXCEL. The core external regions are alternate ring layers of
steel and water up to pressure vessel. The hexagonal cells within
a given radius are automatically identified by the code. The
HEXPIN code has been used for small PWR cores with 7/13/19
assemblies and also for the VVER-1000 MWe reactor core of KK
Project with 163 assemblies. With HEXPIN code the deviation in
power distribution is within 2% of the Russian values in core
interior regions.
Dr. V. Jagannathan, <vjagan@barc.gov.in>
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