23rd Symposium of AER on VVER Reactor Physics and Reactor Safety

Joint AER Working Group A on ?Improvement, extension and validation of
parameterized few-group libraries for VVER-440 and VVER-1000? and AER Working
Group B on ?Core design? twenty second meeting was hosted by ?KODA JS a.s. in Pilsen,
Czech Republic during the period of April 15 to 17, 2013.
In total 16 participants from 5 member organizations took part in the meeting and 11
papers were presented.
Objectives of the meeting of WG A are: Issues connected with spectral calculations
and few-groups libraries preparation, their accuracy and validation.
Presentations were devoted to some aspects of transport and diffusion calculations,
and especially to the benchmark dealing with VVER-440 core periphery power tilt.
G. Hegyi presented paper ?Some final remark on our experience with the
APOLLO2 code?, P. Dalek (co-authors M. Bajan, P. Dalek, R. Zajac and V. Ne?as) spoke
about ?VVER-440 Fuel Assembly Optimization for Uprated Power?, L. Heraltov? provided
?Fullcore VVER-440 benchmark evaluation?, R. Zajac (co-author P. Dalek) informed
about ?Full Core Benchmark Calculations at VUJE?, E. Temesv?ri (co-authors Gy. Hegyi,
G. Hord?sy, A. Kereszt?ri and Cs. Mar?czy) presented ?Solution of the VVER-440
Benchmark by KARATE? and finaly D. Sprinzl (co-authors V. Kr?sl, P. Mikol and J.
?varn?) spoke about "?Full-core? VVER-440 benchmark calculated by MOBY-DICK
macrocode?.
During the meeting were discussed questions connected with the benchmark solution
and, among others, future activities, which are also shortly described in the end of the paper.

DYN3D is a well-known and widely used computer code for reactor physics simulation of
nuclear power plants, in particular for reactors with hexagonal fuel assembly structures. It has
been developed in Helmholtz-Zentrum Dresden-Rossendorf, Germany. The standard version
of the DYN3D code can be used for investigations of transients in light water reactors cores
with hexagonal or quadratic fuel assemblies.
In order to determine the pin with the maximum power in selected assemblies, a twodimensional pin power reconstruction can be performed based on the node homogenized
neutron flux. A superposition of global diffusion solution of the full core calculation with the
assembly pin powers obtained in the cell calculations is used therefore. This method is
implemented in DYN3D for reconstruction of power inside selected assemblies. An improved
onset would be a hybrid solution, the coupling of the full core diffusion solver with an
advanced transport solver on fuel assembly base. This method can be used to directly
determine the power distribution for each rod inside fuel assemblies by applying a transport
solver using unstructured mesh and boundary conditions extracted from the full core diffusion
solution. Nowadays, this mentioned methodology is under development.
In the present work an advanced multigroup transport method of current coupling collision
probability (CCCP) with orthonormal flux expansion inside the calculation regions is being
developed and tested for cylindrical, hexagonal geometries and for assemblies of hexagonal
cells. The results of test calculations demonstrate very good agreement with the results
obtained from Monte Carlo calculations. Multigroup calculations for hexagonal assemblies
with cross-sections prepared using the HELIOS code show good agreement with HELIOS
reference solution, too. These convincing results encourage the implementation of this
advanced pin power calculation method into DYN3D as future pin-power determination
method using currents from nodal solution as boundary conditions.

Two classes of nodal methods were developed with the possibility of the automated mesh
refinement leading to converged solutions. Two benchmark problems were solved for the
verification, namely the static states of AER-2 (VVER-440) and the FCM-101 (VVER-1000)
problem. From practical point of view, the differences between the converged solutions and
those from the CRONOS fine mesh solutions are negligible, nevertheless there are perceptible
deviations. Slicing only the hexagon boundaries for more continuity conditions and parallel
increase the trial functions inside the node gives better performance regarding the computer
time and accuracy than the method of subdividing the hexagon into triangles. Nevertheless,
this later method assures the possibility to take the heterogeneous structure of the node
(burnup, temperature) into account.
1. INTRODUCTION
In spite of the increasing computing performances, coarse mesh nodal methods play an
essential role even nowadays for the tasks where the power distributions of the nuclear reactor
must be determined many times. Core design and optimization or dynamic calculations of the
safety analysis or training simulators are the most important applications in this respect. At
the same time, the accuracy requirements are more and more demanding because the margins
to be applied due to the calculation uncertainties can lead to economic burden in some cases.
These are the reasons why the nodal methods are being developed continuously for a long
time up the present days [1-5].
The KIKO3D code was developed about 15 years ago ⁶ and its nodal method modernization
is going on for three years in accordance with the above reasons. Meanwhile the multigroup
capability was also developed and the new code is called KIKO3DMG ⁷. KIKO3D and
KIKO3DMG are using a special response matrix method for solving not only the steady state
but also the time dependent diffusion equation, two specific additional response matrices are
used for that ⁶. These are the reasons why the methodologies found in the literature could
not be applied directly and a special development was necessary to improve the nodal
treatment. An additional standpoint of the development was that at the same time we wished
to assure the possibility of a mesh refinement up to an arbitrary deep level in order to reach
converged solutions. This later possibility proved useful for determining the uncertainty
originating just from the nodal method. Two classes of the nodal methods were developed. In
the first one the hexagonal nodes can be subdivided to arbitrary number of triangles in an
automated way while in the second one only the node boundaries are subdivided in order to
apply more and more boundary conditions and – in parallel – trial functions. In this latter case,

Submitted paper is devoted to initial value problem for both the neutron transport equation
and the system of equations that describe behaviour of emitters of delayed neutrons.
Examination of the solution to this problem is based on several main suppositions which
concern behaviour of macroscopic effective cross-sections that describe reaction of neutron
with medium, the temperature of medium and the other parameters of the equations.
Formulation of these suppositions is sufficiently general and is in agreement with properties
of all known models of physical quantities mentioned. Among others the suppositions admit a
dependence of both macroscopic effective cross-sections and temperature on spatial
coordinates and time that can be arbitrary to a great extent.
Original mathematical formulation of the problem is the set of integro-differential
equations. At first the problem is transposed into equivalent one to solve certain linear integral
equation for neutron flux. Then solution to that integral equation is found by the method of
successive iterations and its uniqueness is proven.
Numeric solution to the integral equation by Monte Carlo method consists in finding a
functional of the exact solution. In the paper a random process is constructed and several
random variables are proposed to this purpose. Then it is proven that each of these variables is
unbiased estimator of that functional.

ANDREA 2 – improved version of core physics code for VVER
reactors
F. Havluj, R. Vocka
?
?
23rd AER symposium, 30.9-4.10. 2013

DEVELOPMENT OF APPROXIMATION METHOD TO EVALUATE ISOTOPIC
COMPOSITION OF BURNT FUEL
Aleshin S.S, Shcherenko A.I.
Russia, Moscow, NRC ?Kurchatov Institute?
ABSTRUCT
Fuel rods within FA burn out in different conditions in the process of operation due
to the presence of water cavities, absorber and etc. Therefore, the accumulation of isotopes in
different fuel rods goes on in different manner even with identical values of fuel burn-up. The
behavior of isotopic composition can be described in detail by variation spectral code.
However, a full-size core calculation takes much running time, so, it is impractical to realize
the calculation within the framework of design or in-process calculations. In order to take into
account spectral differences during fuel rod operation, it is reasonable to work out an accurate
and fast-operating approximation method to evaluate isotopic composition of burnt fuel.
The paper discusses the method based on spectral functionality – a nonassemptotic
factor which allows to take into account the conditions and history of exposure. The paper
outlines a code based n this method. The method accuracy is determined by comparing the
data by this method against the data obtained by a spectral code.
On this basis of the comparison data, the paper draws a conclusion about the
possibility of using the method to evaluate the isotopic composition of burnt WWER fuel
rods.
INTRODUCTION
The idea of this work was led from (urgency of this work is) the necessity to solve
following problems:
1)
The account of the fissile material in the burnt fuel.
In present time there is planned to repair fuel assemblies by removing the damaged
fuel rods. On this issue we need to have the means to calculate nuclides content in single fuel
rods.
2)
To verify the accuracy of engineering codes.
One way to determine the accuracy of engineering codes is to compare results of
these programs calculation with ones, obtained by the precision software such as MCU, using
the minimum number of approximations (compared with engineering codes).
In that way the main aim of this work is to develop a tool to evaluate the isotopic
content in the burnt fuel rods of WWER reactor core.
There are two different approaches to get the isotopic composition in the fuel rods: to
solve a system of equations for the fuel burn-up cycle or take advantage of approximation
methods.
In this paper, there will be presented a method to obtain the isotopic composition of
the fuel in the cells of WWER using the approximation method that takes into account the
history of fuel depletion by two functional verification – a spectrum nonassemptotic factor and
fuel burn-up.

The optimal choice of mesh points for preparation of table form cross-section library for
DYN3D code is considered in given paper. Mesh points are considered for the following
states of reactor core parameters: fuel temperature, moderator temperature, moderator density
and boric acid concentration. The choice of mesh points and definition of its number is
performed in that way to cover whole possible range of reactor core parameters change
including cold states, hot power states and accident modes. Moreover the best description of
cross section behavior in the interval between mesh points is provided by using the linear
interpolation method in intermediate points.
The work was performed in framework of orders SR 2511 and R0801504 (SR2611)
under German BMU support.

Usually during few-group cross section library preparation for using in WWER macro
calculations the change of nuclide concentration during fuel burnup is defined under
invariable and averaged over whole core characteristics (power, fuel and moderator
temperature, moderator density etc.). In reality conditions of fuel burnup are changing and
this factor effects on fuel nuclide content (so called spectral history effect). For accounting of
real burnup history it?s necessary to take into consideration the dependence of cross sections
on not only burnup, but also on history of neutron spectrum during fuel burnup.
In given report results of analysis of fuel burnup history effect on neutron-physical
characteristics of WWER-1000 core with use of DYN3D code are presented. In DYN3D code
the method of using of the local Pu-239 concentration as an indicator of burnup spectral
history is realized. Calculating studies are performed for the four first fuel loadings of 2nd
unit of Khmelnitsky NPP and stationary fuel loading with TWSA. Effect of fuel burnup
history is presented both on macro characteristics of reactor core (boron acid concentration,
fuel campaign duration, reactivity coefficients) and on local values of burnup and power.
The work was performed in framework of orders SR 2511 and R0801504 (SR2611)
under German BMU support.

MOBY-DICK macrocode historically used finite difference fine-mesh solver to calculate pinby-pin power distribution of the VVER 440 and VVER 1000 reactors. Later a module (called
HanKa) containing nodal solver with pin-by-pin power reconstruction was added and tested.
This module was developed at the Department of Mathematics of the University of West
Bohemia. In this work results calculated using these two solvers were compared. Firstly
results from both calculation types were compared against the data acquired from the ScorpioVVER monitoring system to see how the criteria of the validation test case were satisfied.
Secondly pin-by-pin power distributions from both solvers were directly compared against
each other. It was found that both solvers satisfy the criteria of the validation test case and that
the differences between the two solvers are low enough for the nodal module HanKa with
pin-by-pin power reconstruction being considered suitable for the calculation of the neutronphysical data of the VVER 440 reactor.

The development of the methodological component of engineering factors of the
macrocode MOBY-DICK (MD) ¹ is based on direct comparison of measured against
calculated data. Direct measured data are thermocouple (TC) and Rh detector (SPND)
readings. In this paper the normality of TC versus macrocode deviations is explained and
methodological uncertainties of the macrocode MOBY-DICK is finally checked by 97,5 – th
percentile. Absolute and relative uncertainties of volumetric ( K V ) and radial ( K q ) power
distributions are determined by entire comparisons monitoring power distribution (SCORPIO)
versus MD, TC versus MD and SPD versus MD. The final determination of engineering
factors suppose the hyperbolic dependence of MD methodological uncertainties on power
peaking.

Calculations of neutron-physical constants for software complex “KASKAD” ¹
require the processing of large amounts of information.
This report focuses on the development of a uniform system of NRC KI creating,
editing , recording and storage of all types of data on the fuel assemblies, as well as the
implementation of multi-level access to this data for any number of users.
A GUI is written in C# (C Sharp) programming environment Microsoft Visual Studio
2008. As a server, a relational BANK using Microsoft SQL Server 2008. The GUI combines
an intuitive interface with extensive editing of data, which on the one hand simplifies the
development, on the other hand, allows experienced users to work faster and more efficiently
with a minimum of errors.
The interface provides data processing and editing an object according to pattern, i.e.
opening the fuel assembly is divided into the main components, such as materials, structures
(tubes, tablets), cell (fuel elements, absorbers, burnable absorbers, sensors TDM channels).
Such an approach to the storage and processing of information provides a more systematic
storage of raw data.

In-core power distribution calculated with operative code differs up to some point with
self-power neutron detector readings. Maximum likelihood estimation (MLE) give an
opportunity to receive optimal corrections to operative code results. This estimation give as
power distribution consistent with self-power neutron detector readings, so its most probable
deviation from real power production in the core.
The efficiency of this approach is
investigated. As a real core acts full-scale VVER model based on MCU Monte Carlo code,
which produces power distribution in ~2500 nodes of 3D-mesh.

VERIFICATION OF CODE TDMCC FOR THE NUCLEAR SAFETY ANALYSIS
PROBLEMS FOR VVER-TYPE NPP FUEL SUBSTANTIATION

The ?Full-Core? VVER-440 pin power distribution calculation benchmark was proposed by Mr.
Mikol?? on the 21st Symposium of AER in Dresden in 2011⁴. This benchmark is a 2D
calculation benchmark based on the VVER-440 reactor core cold state geometry with taking into
account the geometry of explicit radial reflector. The main task of the benchmark is to test the
pin by pin power distribution in selected fuel assemblies at the periphery of the VVER-440 core.
In this paper we present our calculations (keff, pin by pin power distribution, integral fission
power) made by MCNP and the KARATE code system in the Centre for Energy Research, the
analysis of our results and some conclusions.

Because of difficulties with experimental validation of power distribution predicted by
macro-code on the pin by pin level we decided to prepare calculation benchmark named
?FULL-CORE? VVER-440. This benchmark is a 2D calculation benchmark based on the
VVER-440 reactor core cold state geometry with taking into account the geometry of explicit
radial reflector. The main task of this benchmark is to test the pin by pin power distribution in
fuel assemblies predicted by macro-codes that are used for neutron-physics calculations
especially for VVER-440 reactors.
The submission of this benchmark was presented on the 21st Symposium of AER in
2011. The reference solution has been calculated by MCNP code using Monte Carlo method
and the results has been published in the AER community. The results of reference calculation
were presented on the 22nd Symposium of AER in 2012.
In this paper we will compare available macro-codes results of this calculation
benchmark.

DYN3D is a three-dimensional nodal diffusion code for steady-state and
transient analyses of light-water reactors with square and hexagonal fuel
assembly geometries as well as of innovative reactor concepts, which is
undergoing continuous development.
In this work, the trigonal-geometry DYN3D diffusion model is verified by
means of the AER-FCM-101 benchmark. With increasing trigonal mesh
refinement, both effective multiplication factor and power distribution
converge well to the extrapolated finite-element reference solution. In
addition, a comparison to the hexagonal DYN3D diffusion models HEXNEM1
and HEXNEM2 is given.

The paper gives an experimental substantiation for the current methods of accounting the
effects produced by the changed shape of VVER-1000 fuel assembly to power distribution in
reactor core. The experimental part of the researches was performed at critical test facility P at
NRC ?Kurchatov Institute?. The neutron-physics calculations of the mathematical model of
the test facility were carried out by computer code MCU-PD. The paper gives in brief the
characteristics of the test facility, description of the types of the discussed compositions of
fuel assemblies, and evaluation of measured and calculated data. We reached a good
agreement between the calculated and measured data that reflect the effects of FA gaps to rod
power density.

STATISTICAL MODEL OF POWER DISTRIBUTION ABOUT VVER-1000 REACTOR
CORE, AND THE MODEL APPLICATION IN THE ANALYSIS OF POWER FIELD
DISTRIBUTIONS IN FUEL CYCLES

FUEL ASSEMBLY BURNUP CALCULATIONS FOR VVER FUEL
ASSEMBLIES WITH THE MONTE CARLO CODE SERPENT
T. L?tsch
T?V S?D Industrie Service GmbH, Energy and Technology (IS-ET), Westendst. 199,
80686 Munich, Germany

CONSOLIDATED DATA AND STATUS OF TASK2 SOLUTIONS OF THE
BENCHMARK FOR CORE BURNUP CALCULATIONS FOR A VVER-1000
REACTOR
T. L?tsch1), V. Khalimonchuk2), A. Kuchin2),
1): T?V S?D Industrie Service GmbH, Energy and Technology (IS-ET), Westendst. 199, 80686
Munich, Germany; 2): State Scientific and Technical Centre for Nuclear and Radiation Safety of
Ukraine (SSTC N&RS), Stusa st. 35-37, 03142 Kyiv, Ukraine

Review of AER Working Group B Meeting in Pilsen, Czech Republic is given.
Regular meeting of Core Design Group was organized by ?KODA JS a.s. and held at Pilsen,
Hotel ?U Pramen??, Czech Republic, April 16?17, 2013, together with Working Group A.
Presented papers (see List of papers and List of participants) covered topics as follows.

The main objective of this research is to develop a new loading pattern designing method
using artificial neural networks and apply it to the Paks Nuclear Power Plant. We expect from
this technique to give us a tool which may allow, combined with the classical refueling
optimization method, a more economical functioning of the reactors of Paks NPP.
In the first part of this paper a short summary can be found about the main characteristics of
the core design at Paks, which is followed by a short introduction into the theory of neural
networks.
The main part of this paper contains the description of our neural network and the developed
computational apparatus. Neural networks are used to predict the value of two target functions
which characterize the safety and the economics of loading patterns. Using the appropriate
network type and structure we can check much more patterns than with the classical
optimization method. From these patterns we can choose the best ones based on the predicted
value of the target functions, the most economical one will be found after the exact
calculation of the function values. This paper contains only the first results of this new
research, we hope in the near future our method can be used as a standard tool for core design.

Pin burn-up optimization by FA?s rotations has been used at Dukovany NPP since 1990s.
This method helps us to uniform pin burn-up distribution and can be very important in the
future with upcoming new fuel designs. The main purpose of this paper is to familiarize
reader with benefits of this kind of optimization. Attention was especially paid to some new
rotation evaluation technique explanation and to show the influence of assembly rotation to
maximal pin burn-up.
1. Introduction
Present day tendency of prolonging of fuel cycles and power increase is connected with
higher pin burn-up in the spent fuel assemblies. Fuel assembly rotation is crucial tool for pin
burn-up optimization.
Under the term ?Assembly rotation? we understand rotation of the assembly by 60, 120, 180,
240 or 300 degrees. Fuel assemblies in VVER-440 core cannot be physically rotated because
of fixation peg and loading machine limitation. We can take advantage of the reactor core
symmetry and rotations can be realized by moving fuel assembly to another sixth of the core
as shown on fig. 1.

Evolution of VVER-440 fuel assembly is characterised by gradual increase of discharged fuel
burn-up and prolongation of fuel cycle. Mentioned effects can be reached effectively by fuel
enrichment raise and by parasitic absorption reduction. Both methods were used during
VVER-440 fuel assembly optimisation, described at this paper. Enrichment interval was
enlarged up to 5.95 % U235, parasitic absorption was limited by carcass frame of the assembly.
Basic characteristics of the best assembly proposals are presented and effects on equilibrium
fuel cycle of VVER-440 reactor are characterised.

23rd SYMPOSIUM of AER
on VVER Reactor Physics and Reactor Safety
September 30 – October 4, 2013, Strbske pleso, Slovakia
PRINCIPAL RESULTS OF EXPERIMENTAL ?COMMERCIAL OPERATION OF
VVER-440 FUEL OF 3-D GENERATION
V. V. Saprykin. A. A. Gagarinsky
NRC Kurchatov Institute, Moscow
V.Adeev-Kola NPP,
Russia
1. Introduction
On the Kola NPP, Unit 4 trial operation of the 3-rd generation fuel continues. This fuel have a
number of design features, providing the best operational characteristics [1, 2]. Increasing of
efficiency of nuclear fuel usage will be achieved by reduction of the parasitic capture of thermal
neutrons in constructional materials (weight of zirconium is reduced), optimization of uranium
water relation (increase in fuel elements step), increasing of uranium loading (usage of fuel
pallets with increased diameter and without central hole in them).
The basic characteristics of the core with fuel of the 3rd generation are provided in work
³. In the present report, being addition to the report ⁴, gave new results of operation and the
short analysis of the obtained data is made. Experimental characteristics of WWER-440 reactor
core with fuel assemblies of the 2nd (FA-2) and 3rd (FA-3) generation of the enrichment
increased to 4.87% are submitted. Some questions of operation of FA-2 and FA-3 are discussed:
assessment of influence of cover absence on indications of thermocouples at joint operation of
FA-2 and FA-3, features and methods of core design.
2. Purposes of trial operation.
Trial Operation is carried out with the purpose of confirmation of operability FA-3 during 7 fuel
cycles. Use of only 12 RK-3 is planned. Placement of RK-3 by symmetrically RK-2 of identical
enrichment, for comparison of their characteristics for various positions in core is supposed. As
result of trial operation using of these FA in all WWER-440 reactors will possible.
3. Comparison results of experimental and calculated data on the energy release.
In the first fuel cycle FA-3 were established on the core periphery in 3 pieces conjointly in 4
sectors 60 of grad symmetry. In the remained 2 sectors FA-2 of 4.87% enrichment are located. In
the following fuel cycles placement of FA-3 and FA-2 in other characteristic positions of core is
supposed. In the 3-th fuel cycle FA-3 were placed near to working group of the control and
protect system.
For determination of true power of FA-3 by the temperature measurements the great
value of influence of intra- and inter-assembly leakages of coolant can have. The assessment of
influence of lack of a assembly cover on indications of thermocouples is important for safe
operation since basic data of core there are indications of the temperature sensors located at the
exit from FA.
Calculations by the SC-1⁵ program have shown that for the lack of assembly?s cover a
coolant flow which can participate in removal of heat of FA-3 is bigger than corresponding flow
for FA-2 on 10% ⁶. In conditions of absence of assembly?s cover, mixed value of coolant
temperature on going out of fuel bundle characterized FA-3 power only approximately. Design
features of FA-3 define nature of change of pin by pin energy release in these assemblies.

Paks NPP status and transient to 15 m cycles
Imre NEMES

Different kinds of Gd-fuel have been used at Dukovany NPP since the year 2003. All of them
were designated for 5-year fuel cycle strategy utilisation. Implementation of such strategy
needed to modify approaches to fuel cycles planning and fuel utilisation efficiency evaluation.
Chosen method of this particular problem solution is explained and described at this paper.
Attention was paid to assemblies and fuel pins burnup evaluation especially.

AER WORKING GROUP C ACTIVITY IN 2013

The SCORPIO-VVER core monitoring system has proved since the first installation at
Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It
is now installed on four units of Dukovany NPP (Czech Republic), on two units of Bohunice
NPP (Slovak Republic) replacing the original Russian VK3 system and on the full scale plant
training simulator at the Centre for training and education of the reactor operators and reactor
physicist in Trnava (Slovak Republic). By both Czech and Slovak nuclear regulatory bodies
the system was licensed as a Technical Specification Surveillance tool.
Since it?s first installation, the development of SCORPIO-VVER system continues along with
the changes in VVER reactors operation. The system is being adapted according the utility
needs and several notable improvements in physical modules of the system were introduced.
The latest most significant changes were done in connection with implementation of a new
digital I&C system, loading of the optimized Gadolinium bearing Gd2 fuel assemblies,
improvements in the area of core design (neutron physics, core thermal hydraulics and fuel
thermal mechanics), adaptation of the system to up-rated unit conditions, in design and
methodology of the limit and technical specifications checking (implementation of the on-line
shutdown margin calculation to the system) and improvements in the predictive part of the
system (Strategy Generator).
After the currently finished upgrades the SCORPIO-VVER is still in focus of Central
European nuclear power plants with the roadmap of modification and implementation up to
2016.
In October 2012 the Upgrade 3 for the support of the new type of fuel with 4.87% enrichment
of 235U loaded into the reactors at EBO Slovakia has been finished successfully. Thanks to
the upgraded system all units could reach their 100% of nominal power.
During June 2013 the 6th Upgrade was started at EDU NPP in Czech Republic focused to the
SCORPIO-VVER enhancement and new function development up to 2016.
The Dukovany?s planned Upgrade 6 presents the most advanced upgrade in the system?s
history. The SCORPIO-VVER system will be completely renewed, re-hosted to the new
hardware, implemented the latest advanced calculation codes with enhanced accuracy and
adapted to the local, national and international IEC standards and requirements too.
With the upgrade on the way and new installation under the preparation, the SCORPIOVVER project is very active and prepared to meet new challenges in future.

The paper follows the similar paper ?Primary system hydraulic characteristics after
modification of reactor coolant pumps` impeller wheels at Bohunice NPP? presented last year
in AER Symposium 2012 in Pr?honice [ 1 ].
The project ?Modification of reactor coolant pumps impeller wheels at EBO NPP? is
continuing also in 2013. During the outages in 2013 two impeller wheels were replaced at
both units. Nowadays Unit 4 is operated with all 6 new impeller wheels and Unit 3 with four
new impeller wheels. Modification of last two impeller wheels at Unit 3 will be performed
during the outage in 2014.
The paper presents the main results of non-standard primary system hydraulic
characteristics measurement executed in 2013. Comparison with expected values is presented
as well. Measurements showed that the increase of coolant flow through the reactor is in good
agreement with expected values. The increase of the coolant flow through the reactor was
proved not only by primary system hydraulic characteristics measurements, but also by
thermal balance measurements. Both measurements are based on different unit parameters.
The increase of coolant flow through the reactor in comparison with initial value in 2011 is at
Unit 4 circa 4,5 % (after modification of all 6 impeller wheels) and at Unit 3 circa 3,2 % (after
modification of 4 impeller wheels).

The paper demonstrates the process of asymptotic spectrum formation in the course of
neutron power decline after the negative reactivity was inserted in VVER reactors. The
interest in this subject is caused by the requirement of making calculated corrections in order
to condition the measured data on the worth of VVER CPS control rods.

The AER Working Group D on VVER reactor safety analysis held its 22th meeting in Paris,
France, during the period April 10-11, 2013. The meeting was hosted by the OECD/NEA and
was held in conjunction with the fifth workshop on the OECD/NEA Benchmark for the
Kalinin-3 VVER-1000 NPP and the seventh workshop on the OECD Benchmark for
Uncertainty Analysis in Best-Estimate Modeling (UAM) for Design, Operation and Safety
Analysis of LWRs. Altogether 17 participants attended the meeting of the working group D,
12 from AER member organizations and 5 guests from non-member organization. The coordinator of the working group, Mr. S. Kliem, served as chairman of the meeting.
The meeting started with a general information exchange about the recent activities in the
participating organizations.
The given presentations and the discussions can be attributed to the following topics:
?
?
?

The paper proposes the techniques for evaluating emergency protection effectiveness
with a stuck most effective CR using the measured worth of certain CR groups. The paper
gives estimation of probable uncertainties of weight coefficients, which are introduced to the
worth of individual CR group in EP weight, and the proposed techniques as a whole.

The fulfillment of the safety analysis acceptance criteria and the normal operation safety
related limitations are usually evaluated by using separate hot channel or/and hot assembly
thermal hydraulic calculations. In accordance with the present structure of the VVER-440
fuel assemblies, the subchannel thermal hydraulic calculation of the recent KARATE-440
code system exploits the presence of the assembly shroud by applying zero mixing
between the assemblies. However, the shroud-less assemblies of the expected new NPP
units in Hungary or even those of some recently foreseen VVER-440 are raising several
questions; like what is the role and the influence of the mixing from or/and to the
neighboring assemblies, and in this new situation which is the smallest possible calculation
domain in the vicinity of the hot channel. Focusing on ATWS problems, the paper
evaluates the thermal hydraulic influence of the environment of the hottest subchannel by
supplementing the calculation domain step by step with the closest pins and subchannels of
the neighboring assemblies. The pin power distributions are also varied. The conservatism
of the closed assembly approximation for the shroud-less case, neglecting the hydraulic
mixing effect between the assemblies, is also analyzed.

The thermohydraulic computer code KORSAR is referred to the best estimate codes and
intended to simulate the design and accident conditions of VVER reactor plants. To perform
the coupled neutron-physical and thermal-hydraulic calculations using this computer code the
Code package (CP) SAPFIR_95&RC_VVER is used and by means of the latter the constant
support of program block KARTA, is prepared for the calculation of neutron kinetics of
VVER-type reactor cores in 3-D approximation.

Nowadays, there is a tendency to use best estimate plus uncertainty methods in the field of
nuclear energy. This implies the application of best estimate code systems and the
determination of the corresponding uncertainties. For the latter one an OECD NEA
benchmark was set up [1, 2]. The objective of the OECD NEA Uncertainty Analysis in BestEstimate Modeling (UAM) LWR benchmark is to determine the uncertainties of the coupled
reactor physics/thermal hydraulics LWR calculations at all stages.
In this paper the focus is on the determination of the uncertainties of the thermal-hydraulic
analyses of rod bundles. For this purpose a 5*5 bundle with electrically heated rods
representing a 17*17 PWR bundle was selected for these investigations.
In the first part of the paper the modeling uncertainties are evaluated using the OECD
NEA/NRC ³ PSBT benchmark data. The main goal of this section is to determine an
appropriate turbulent mixing factor within its uncertainty for this type of bundle.
After that the uncertainties of the COBRA calculations are discussed using Monte-Carlo type
statistical analyses taking into account the modeling uncertainties and other uncertainties
prescribed in the UAM benchmark specification: e.g. geometrical (diameters, rod pitch)
uncertainties, boundary condition uncertainties like uncertainties of pressure, power, etc. Both
steady-state and transient cases were investigated. The target quantities are the uncertainties
of the void distribution, the moderator density, the moderator temperature and the DNBR. In
many cases both bundle averaged and single sub-channel results are presented.
We will see that – beyond the uncertainties of the geometry and the boundary conditions – it is
very important to take into account the modeling uncertainties in case of bundle or subchannel thermo-hydraulic calculations. Another important and interesting conclusion is the
strong direction-dependence of the target uncertainties in the axial direction.

An uncertainty analysis of a large break loss-of-coolant accident (LBLOCA) for a German PWR
Konvoi was performed using a statistical method, which is based on the Wilks? theory. The
evaluated output parameter is the peak cladding temperature (PCT). The primary goal of this
paper is a ranking of the input uncertainties, according to their contributions to the PCT
uncertainty in the ATHLET simulation of PWR LBLOCA, by performing a sensitivity analysis.
It was shown, that the starting (extended) set of varied parameters can be considerably reduced
without any statistically significant influence on the uncertainty analysis results.
The statistical t-test was used to minimize the number of varied parameters. A set of uncertainty
parameters with significant impact on the uncertainty of the PCT was identified. The main
contribution to the uncertainty of the first cladding temperature maximum during the blowdown
phase of the accident is produced by the core parameters that affect the fuel?s stored energy at
the beginning of the accident. However, the major contributors to the uncertainty of the second
PCT maximum are the uncertainties in the code models, and first of all the uncertainties in the
heat transfer coefficients for dispersed and pure steam flows.
1

In the paper is represented the robust method of calculation of reducing engineering
factor for DNBR, which can be used in ?hot channel? model in nuclear reactor safety analysis.
The estimation of DNBR producing with help of method of statistical account of
uncertainties. The method is based on the special developed FA model which includes
mathematical models of energy field calculations, FA thermomechanical deformations,
thermohydraulics and technological processes. FA thermohydraulics calculations carry out
using cell by cell thermohydraulic code SC-1 ¹. In these calculations were taken into
account deviations in different parameters.
Due to construction distinctions of different type cassettes the values of DNBR and
engineering coefficients in these cassettes can be vastly different. In the paper is shown
comparison of DNBR values and reducing engineering factors for DNBR in TVS-2M and
TVSA. The sensitivity of reducing engineering factor for DNBR was analyzed for different
factors: changing of power level, relative axial heat flux profiles, rods energy distribution, the
law of statistical distribution random parameters, deformation of energy field.
The researching of stability minimal DNBR values localization in normal and
emergency processes was done.

The present paper is a continuation of the uncertainty and sensitivity (U&S) study of the
VVER-1000 reactor, already started in [1-2]. It is based on the OECD/NEA coolant transient
Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP) where a switch
off of one main coolant pump (MCP) at nominal reactor power is analyzed ³. The novelty of
the work consists of taking into consideration of all hydraulic uncertainty parameters used in
the modeling of the reactor pressure vessel (RPV) internals. A detailed parallel channel
ATHET model of the RPV is developed. It consists of 26600 control volumes most of them
connected with junctions for cross flows. The specific geometry of the gap between upper
part of the baffle and upper part of fuel assembly and also a fuel assembly head are taken
explicitly into account. To reach the specified goals the GRS methodology for uncertainty and
sensitivity analysis ?SUSA? is applied. According to it, 500 samples are generated taking into
account 50 hydraulic parameters in the ATHLET RPV model. The ranking of uncertainty
parameters is obtained and analyzed in detail. The influence of the input parameters on the
uncertainties of the assembly-wise mass flow and coolant temperature axial distributions is
shown.
The significance of the paper consists in pushing forward an approach of ?best-estimate
calculations with a validated system code in combination with performing uncertainty and
sensitivity analysis?. The performed work is the first step in the application of the global
uncertainty parameters? study for K-3 benchmark. Parallel to this work studies are launched
on the influence of nuclear data uncertainties and thermal parameters on the final results.

The paper presents a methodology which will enable in the future performing of fast transient
analysis of NPP with VVER-1000 reactors. It is based on artificial neural networks (ANN)
method. For the training of the network the best-estimate system code ATHLET (GRS) is
used with detailed nodalization (multi-channel model) of the Reactor Pressure Vessel (RPV).
The present work is dedicated to the training procedure which is connected at this stage of
development only with the setting in the system the RPV hydraulic resistances. Training
samples are created by random variation from a double sided 95% confidence interval of 50
different hydraulic resistance coefficients in the RPV. For each set of hydraulic resistance
coefficients a test calculation is performed with ATHLET to obtain the corresponding thermohydraulic distributions (mass flow, pressure, temperature) within the RPV. For this training
procedure the total and the assembly power distributions used in all calculations remain
unvaried and equal the nominal values, e.g. at this stage a dynamic 3D neutron-physics model
is not being taken into account. As a result of the neural network training an application
program is created. This program is capable to give almost simultaneous answer about all
thermo-hydraulic fields at each point of the RPV. A comparison of the system results with
results obtained from the AHTLET simulation gives a good correspondence for such nodes
(fluid objects) of the active core, which have not been taken into account in the training
sequence of the artificial neural network. The paper contains also some ideas how to apply the
artificial network modeling of the RPV in the current practice.
This work is a step forward to create an artificial neural network system for performing of fast
NPP analysis. Further training of the system will be done with the help of coupled neutronphysics/thermal-hydraulics codes based on ATHLET.

Analysis Compare of Uncontrolled Withdrawal of Control Rods Group
at a Certain Power of VVER and M310
WANG Yanping, ZHAO Yang, DU Sijia
National Key Laboratory of Science and Technology on Reactor System Design Technology
Nuclear Power Institute of China, Chengdu, 610041
(yizimu0005@gmail.com, +86-28-85908959)

VUJE, Inc.
Okru?n? 5, 918 64 Trnava
Slovensk? Republika

An important issue of safety considerations for nuclear power plants is the determination of
the necessary submergence of pump intakes to ensure a reliable pump operation during postulated abnormal operation and accidents. Surface vortices occur when the water level above
the pump intake is decreased below a critical value. These vortices induce swirling flow and
gas entrainment at the intake which lead to the decline of flow or even pump failures and
damages. Thus, surface vortices must be considered for the determination of the available
amount of cooling water in the reservoir tanks and reactor sump of nuclear power plants. The
prediction of the critical submergence is difficult, because it depends on many geometrical
and fluid dynamical parameters. The critical submergence is generally calculated with universal and conservative easy-to-use correlations, if experimental data is not available. However,
these correlations are only valid within a limited parameter range. An alternative, more detailed approach to calculate the critical submergence is investigated within the research
project ?Generic numerical determination of the critical submergence at pump intakes to
avoid gas entrainment? sponsored by the German Federal Ministry of Economics and Technology. The developed calculation approach is based on the combination of ANSYS CFX
with the analytical vortex model of Burgers & Rott. This paper describes the application and
validation of the improved method against the experiments from Jain et al. Additionally the
applicability and the limitations of the widely-used correlation of the American National
Standards Institute (ANSI) for the calculation of the critical submergence have been investigated by using the improved method.

In nuclear energy, continuous developments aim for safer operation and higher efficiency.
In order to achieve these goals, detailed knowledge of thermal hydraulics of fuel assemblies has
high importance. The advanced and properly used CFD (Computational Fluid Dynamics) codes
can give detailed information on thermal hydraulic processes in fuel assemblies. In this paper
steps of the calculations for the bare rod bundle and so-called swirl type spacer grid are showed
and the effects of choice of model details are examined. The modeled pin bundle and spacer grid
geometry is related to rectangular western type PWR fuel assemblies.
First a CFD model has been developed for a subchannel in order to carry out a mesh
independence study. Based on the results a model for the half cross section of the test bundle has
been developed and calculations are carried out with different turbulence models. Translational
periodicity is applied in axial direction in order to calculate a fully developed flow, which will be
used as inlet boundary conditions for spacer grid calculations. Results are compared with LDA
(Laser Doppler Anemometry) measurements published in the frame of the OECD NEA MATISH benchmark. As a last step a model for the swirl type spacer grid is developed and calculations
are carried out with different turbulence models using the results of the bare rod bundle
simulations as inlet boundary condition.

INVESTIGATIONS OF AXIAL DISTRIBUTION OF FUEL ROD SIMULATOR
TEMPERATURE PERFORMED AT A 37-ROD FA MODEL OF REACTOR
VVER-1000

AER Working Group E in 2013
Vladim?r Chrap?iak, VUJE, Inc., Slovakia, vladimir.chrapciak@vuje.sk

The speci?cation of CB6 benchmark on VVER-440 ?nal disposal was published by OECD/NEA
Working Party on Nuclear Criticality Safety (WPNCS), Expert Group on Burn-up Credit Criticality Safety (EGBUC). The benchmark is part of the assignment of EGBUC in assessing the
ability of code systems to predict the reactivity of spent nuclear fuel systems.
The CB6 benchmark is divided in two parts. The ?rst part is aimed at investigating the evolution
of the isotopic composition of an irradiated VVER-440 fuel assembly over the time duration
relevant to long-term spent fuel disposal. The initial isotopic inventory for 4.25 % VVER-440
fuel at 50 MWd/kgU burn-up is given in the benchmark speci?cation.
The second part of the benchmark involves calculation of e?ective multiplication coe?cient
values for 3D cask geometry at various time points. The cask is loaded with 84 VVER-440 fuel
assemblies and ?ooded with pure water. Isotopic composition of the fuel, which is assumed to
be uniform, results from the decay calculation. The “actinides and ?ssion products” burn-up
credit approach is applied in the calculation.
In this study, the code Serpent is applied in solution of the benchmark. Serpent is a continuousenergy Monte Carlo reactor physics code developed at VTT Technical Research Center of Finland. In order to gain comparison data, the decay calculation part is also carried out with
point-depletion calculation code ORIGEN.
The paper demonstrates the decay and ke? calculation results as determined in the benchmark
speci?cation. The di?erences in results arising between the two calculation codes in decay
calculation, and between the ENDF/B-VII and JEFF-3.1.1 based cross section libraries in ke?
calculation, are depicted and discussed.

The WIMS9a ⁴ (and WIMS10a (beta4) ⁵) reactor-physics computer programs
were used to simulate the burn-up of UO2 fuel in a VVER-440 reactor.
The manner in which WIMS9 (and WIMS10) were used in these simulations is
described.
Predictions of concentrations for actinides up to Cm244 and for selected fission
products were compared with laboratory-measured values.
Reasonable agreement between calculated and measured values was seen for some
isotopes, but also differences were observed for the others. Also, agreement is very poor for
some isotopes in selected samples.
Differences between calculated values by WIMS9 and WIMS10 (with different
library) and WIMS10 – 2D standard model vs. WIMS10 3D model are also shown.

Preliminary decommissioning plans constitute vital parts of the licensing
documentation of nuclear power plants. On the other hand, a reasonable estimate of the
amount and radiological characteristics of the radioactive waste born during the operational
life of a power plant is crucial. Since the largest activities are created by neutron activation,
the characteristics of the activated waste materials, such as mass, volume, specific activities
by isotopes and waste index, should be estimated in advance as accurately as possible in order
to assess the waste repository requirements of an NPP. The cost issues related to waste
disposal are huge. Therefore, inadequate planning of the volume and mass of low,
intermediate and high level waste types can result in financial problems.
The present paper describes the methodology applied for determination of neutron
induced activation of steel components and concrete structures surrounding VVER 440 type
reactor cores and the general results which were obtained from the above mentioned two
projects. According to related international literature and studies performed earlier by the
authors, the vast majority (usually exceeding 99%) of the radioactive inventory present at the
time of final shutdown of a nuclear power plant is concentrated in the materials directly
surrounding the reactor core. In these regions, the predominant evolution of radionuclides is
upon neutron induced nuclear reactions. This report is not aimed at the assessment of the
radioactive inventory produced by contamination of NPP components (such as steam
generators and pipelines etc.), only the relevant radioactive waste characteristics of the
activated steel and concrete structures are calculated and summarized.

The SCALE system is very well known and is used for spent fuel calculations – criticality and
inventory. The latest version is the SCALE 6.1.2 system.
By post Fukushima criticality analyses we have used very simple model of assembly by
calculation of losing of cooling water. In this article are compared simple and very detailed
model of assembly and different water density.
By inventory calculation with the SCALE 6.1.2 system is possible to use several different
modules: ORIGEN-S, TRITON + NEWT (2D model) or TRITON + KENO VI (3D model).
In this article are compared activity, decay heat, neutron and gamma source for 1 assembly by
using different modules.

Fuel depletion calculation is computationally expensive. Deterministic transport code requires
large amount of computational time that can be cut down by decreasing the number of energy
groups. Current multigroup energy structures with about 200 groups were generated
by merging broader structures and subsequent ad hoc adding of energy boundaries. Recently,
TSUNAMI sensitivity and uncertainty code sequence from SCALE code package was
developed and released. TSUNAMI allows sensitivity calculations that can be used in various
application areas, one of them is energy structure improvement and development.
TSUNAMI code sequence was used to generate sensitivity coefficients describing sensitivity
of multiplication coefficient to neutron energy. Based on sensitivity coefficient values during
depletion, importance of energy groups was defined. Initial 238-group energy structure was
collapsed to new one according to importance data. Fuel depletion calculation with TRITON
code sequence from SCALE code package was executed and multiplication coefficient during
depletion was compared. VVER-1000 fuel assembly with Eu and Er burnable absorbers were
analyzed.

The paper briefly describes the issue of radionuclide inventory determination in
decommissioned nuclear power plants with emphasis on the VVER-440 reactor type.
According to radionuclide inventory calculation, it is possible to optimize the time frame and
choose the appropriate dismantling procedure during the disposal of reactor internal and
external components during the decommissioning phase of a nuclear power plant. Due to high
technological and financial demands in the case of in situ measurements or sampling, it is
important to use validated code systems to precisely estimate the radionuclide inventory.
Prerequisite for a correct calculation is the collection of reactor operation data. In this paper,
abilities of MCNPX code in this field is presented on the example of calculation for the
shielding (dummy) elements used in V-230 reactor in Jaslovsk? Bohunice Nuclear Power
Plant. These elements were installed into the reactor core in order to protect the reactor
pressure vessel from high neutron flux. They also represent the most activated component of
the reactor.

Review of AER Working Group F Meeting in Moscow, Russia is given.
Regular meeting of 14th session of the Spent Fuel Transmutations Group was organized by
National Research Centre ?Kurchatov Institute? and held in Moscow, Russian Federation, July
15 ? 18, 2013. Presented papers (see List of papers) covered topics as follows.
It is known that conversion of nuclear energy to self-sufficient fuel supply in the second half of
the XXI century is planned to achieve by nuclear fuel cycle closing, intensive deployment of fast
reactors (FR) and improving of fuel usage efficiency in thermal reactors (TR). Only by closing
the nuclear fuel cycle and involving of fast reactors, nuclear power acquires properties that
enable it to assert itself as a power technology of the future: a) the energy potential of nuclear
power is much greater than all the reserves of fossil fuels, b) the amount of long-lived
radioactive waste is equal not to the time integral of the burnup, but only proportional to the
energy capacity of the system.
Apart from the introduction of fast reactors to the nuclear power structure, it takes at least one
more reactor type with the main function of transmutation of all long-lived actinides and some of
the most dangerous fission products produced by nuclear power system. It is a reactor, that
should provide ultimate closure of the nuclear fuel cycle.
The delay in the development of the required FR fraction, producing thermal reactors fuel in the
nuclear power system, will result in issues with their fuel supply by the mid-21st century. This
can be a significant factor in deterring the development of nuclear energy.
In this regard, in terms of the main areas for research and development on controlled
thermonuclear fusion (CTF) it is appropriate to include those that are focused on solving urgent
nuclear power problems of foreseeable future, challenges that are already quite clearly marked. It
is unlikely that this can delay the beginning of the thermonuclear age, quite the contrary, it can
stimulate the development of the CTF, if it is shown in practice its high efficiency (or even
uniqueness) to resolve these problems. Nuclear Energy of the XXI century based on fission
reactors with thermal neutrons and fission reactors on fast neutrons to be developed should have
practically unlimited fuel resources. In connection with these ideas the development of fusion
neutron sources with the main purpose of production of nuclear fuel for fission reactors is urgent.
Preliminary system evaluation shows that the STN (Source of Thermonuclear Neutrons) fraction
with the main purpose of thermal fission reactor fuel production can reach in terms of energy ~
(7-10 )% of the nuclear power capacity, when the rest ~ ( 93-90 )% of the capacity of nuclear
power system will be produced by thermal fission reactors, provided with fuel from STN
thorium blankets. In this case, the benefits of nuclear fission systems effectively complemented
1

The European version of Supercritical Water Reactors (SCWR), the High Performance Light
Water Reactor (HPLWR) operates in the thermodynamically supercritical region of water. Our
basic objective was to elaborate a stationary coupled neutronic-thermohydraulic code capable for
the calculation of the actual 3-pass core design with fuel assembly clusters. The calculations
covered the neutronic transport calculations of HPLWR fuel assemblies, the coupled neutronicthermohydraulic global calculations and the pin-wise analysis. Applying conservative
assumptions, the relation to the linear heat rate and maximum cladding temperature limits was
checked for the equilibrium cycle of HPLWR with this new code system.

ALLIANCE
for
ALLEGRO
Branislav Hatala
Petr Da??lek
23rd Symposium of AER
on VVER Reactor Physics
and Reactor Safety
?trbsk? Pleso, Slovakia,
October 30 – November 4, 2013
www.vuje.sk

The first of all this report represents the formed analytical techno-economic model of
integrated quantitative analysis of (local) performance and operating characteristics of a
distributed system consisting of production units (nodes) of the closed fuel cycle nuclear
power. The proposed simulation model takes into account all the important process blocks of
two conjugate cascade of nuclear reactors at the “fast – thermal” neutrons. The structure of the
simulation model for each stage of a typical nuclear power plant is taken into account
individually distributed structure of the interacting of process nodes (units): 1) the production
of fuels (fuel elements), 2) energy production by nuclear power reactor facilities of standard,
as energy producers fissile isotopes (U-235, U -238, Pu-239, Th-232), 3) regeneration of
spent nuclear fuel materials, 4) the management of radioactive waste, 5) decommissioning of
nuclear power block structures. The corresponding structural and complex system is a closed
multi-level (hierarchical) information-analytical model with multiple feedback loops, taking
into account information about the deterministic estimates of parameters (performance status)
network of production nodes.