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KUDANKULAM 1000 MWe VVER

128 BARC HIGHLIGHTS Reactor Technology & Engineering

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 1 0 0 0 M W e V V E R

I N T R O D U C T I O N

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.


129 Reactor Technology & Engineering BARC HIGHLIGHTS

V V E R is an acronym for “Voda Voda Energo Reactor” meaning

water-cooled, water moderated energy reactor. The VVER

reactors belong to the family of the Pressurised Water Reactors

(PWRs). The KK-VVER has a three-year fuel cycle. This reactor

requires annual refueling of one third of the core i.e.,

approximately 55 fuel assemblies.

The reactor plant consists of four circulating loops and

a pressurising system connected to the reactor with each loop

containing a horizontal steam generator, a main circulating pump

and passive part of emergency core cooling system

(accumulators). The loops are connected with the reactor

pressure vessel assembly by interconnected piping. The reactor

also consists of a reactor protection and regulation system,

engineered safety features, auxiliary system, fuel handling and

storage system.

Major Systems of 1000 MWe VVER



6.1 VVER-1000 MWe REACTOR ANALYSIS

The VVER-1000 MWe reactor core of Kudankulam (KK) Project

is a Pressurized Water Reactor (PWR) of Russian design. It is

necessary to develop indigenous capability of in-core fuel

management of these reactors. This capability is also essential

for an in-depth review of the PSAR documents submitted by

Russian Federation for KK Project. The detailed analysis and

comparison of results with the Russian design project reports

giving the physical characteristics under various steady state

conditions has revealed that it is essential to develop capability

for analysing some of the slow (xenon) and fast transients.

Indigenous lattice burnup code EXCEL and core diffusion

analysis codes TRIHEX-FA and pin-by-pin simulation code HEXPIN

have been developed and are used to analyze the KK core.

After generating the complete lattice database with EXCEL code

for 11 fuel types, the VVER-1000 Mwe reactor core of KK Project

was followed up for 8 fuel cycles. Each hexagonal assembly cell

was divided into 54 triangular meshes. The results like critical

soluble boron, radial and axial power distribution, 2-D and 3-D

peaking factors were compared with Russian data. The calculated

critical boron with a uniform keff normalization agreed well with

Russian data for all eight fuel cycles. The deviation was slightly

more for first fuel cycle, possibly due to non-equilibrium Sm

load. Power dependent feedback is being implemented

in TRIHEX-FA code to reduce the deviations observed in power

distribution. The modeling of reflector region is also being fine

tuned.

Typical Comparison of Radial Power and Burnup Distribution Core Follow-up Simulation for KK Cycle-8

V. Jagannathan, <[email protected]>



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 analys is of physics

commissioning experiments and several reactor-

years of operat ional exper ience 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 val idation for the modern BWR, PWR,

VVER and PHWR. Though the computat ional

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 effect ive 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 distr ibution and hot

spots wi l l provide a good ins ight into the

analysis and also enable intercomparison of different

nuclear datasets and methods. S ince 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



VISWAM: Present Status of Code development

Typical Screen View of ‘XnWlup’ Software toView Microscopic Cross sections

Typical Screen View for Input for VVERAssembly Cell Description

The visual mode of creating input to typical VVER fuel assembly

description is also given. The visual input can be internally

transferred to the other digital form of input.Typical viewing of multigroup cross section by the ‘XnWlup’

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.



V. Jagannathan, <[email protected]>

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 Epithermal Flux Profile inTAPS BWR – by Plot3D

Typical Thermal Flux Profile in a ThoriumLoaded Reactor – by Plot3D

Typical Thermal Flux Profile in a LWR

Typical Screen View of Input for VVERAssembly Geometry



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, <[email protected]>

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