21st Symposium of AER on VVER Reactor Physics and Reactor Safety

21st Symposium of AER on
VVER Reactor Physics and Reactor Safety
Dresden, September 19 – 23, 2011
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21st Symposium of AER on
VVER Reactor Physics and Reactor Safety
Dresden, September 19 – 23, 2011
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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? twentieth meeting was hosted by ?KODA JS a.s. in Plze? (Czech
Republic) during the period of 9th to 11th May 2011.
In total 19 participants from 7 member organizations were present 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
also benchmarks dealing with VVER-1000 and VVER-440 core periphery power tilt.
Zs. Sz?cs?nyi (co-author I. Nemes) presented ?Investigation of highly enriched fuel
for VVER-440 application using different geometry and enrichment profile?, R. Zajac (coauthors P. Dalek, C. Strmensk? and J. Majerk) spoke about ?VVER-440 with IMF Safety
Analysis of Control Rod Ejection?, G. Hord?sy (co-authors Gy. Hegyi, Cs. Mar?czy and E.
Temesv?ri) presented paper called ?Preliminary Solution of the ?MIDICORE? VVER-1000
Core Periphery Power Distribution Benchmark by KARATE and MCNP?, P. Dalek (coauthors M. Antal and D. Bun?ek) spoke about ?First utilization of GD2 fuel with enrichment
4,87% at SE-EBO?, P. Mikol (co-author P. Vesel?) presented paper called ?VVER-440
Fuel Cycles with Increased Enrichment and Uranium Mass? and D. Sprinzl (co-authors V.
Kr?sl, P. Mikol?? and J. ?varn?) provided a ?Full-Core? VVER-440 pin power distribution
calculation benchmark proposal?.
Future activities are also shortly described in the end of the paper.

In the early years, after the discovery of fission, the criticality safety assessment and the
established safety limits, have been mainly based on direct experiments. Later, following the
advances in the theory, computational methods and computer hardware, theoretical methods
have been elaborated to the level to become reliable assessment tools. The computer codes
started replacing the experiments, while the experimental data became a valuable validation
source for their models.
An application of the 2D transport theory code HELIOS (1,2) for assessment of criticality
limits is presented in this paper. The effect of the enrichment, the system dimensions, H/U5
ration and different reflectors were studied in heterogeneous and homogenized systems.
Comparisons with published experimental data and evaluated safety limits are made here to
demonstrate the range of HELIOS applicability and limitations.

The full-scale three-dimensional computer model describing the physical processes
that occur in the reactor core of the VVER-1200 assumes the detailed description of the
geometry of the FA. However, detailed description of the geometry essentially complicates
the tasking of the entrance data, increases the time of calculation and probability of errors. As
a consequence, creating the simplified models of the design elements becomes necessary. In
the present work the simplified geometrical models of FA?s design elements are offered.
Comparative calculations of the simplified models and the exact models completely
corresponding to the working drawings of a general view ?KB “Hydropress” are carried out.
It is shown that in many problems using of simplified description of the FA elements
geometry is possible.

This paper is dedicated to the latest version of TVS-M code ? TVS-M 2007, which
allows the neutron flux distribution inside fuel assemblies to be calculated without using the
diffusion approximation. The new spatial calculation module PERST introduced in TBS-M
code is based on the first collisions probability (FCP) method and allows the scattering
anisotropy to be accounted for.
This paper presents some preliminary results calculated with the use of the new
version of TVS-M code.

A new analytical method is described to deal with the Leakage Environmental Effect. The
method is based on the analytical solution of the two group diffusion equation for two adjacent
fuel assemblies avoiding the eigenvalue iteration by introduction of an external source. The
quality of the results for this highly efficient method is demonstrated for quadratic fuel
assemblies. In additional tests the transferability of the concept to hexagonal VVER-440 type
fuel assemblies is shown.

VVER core is located inside a metal core basket. There is a possibility of the some
FAs displacements within a basket in the transverse direction in the design tolerances, as a
result of assembly bending or during mounting of the core. We have investigated several
different variants of changes in the gap size between fuel assemblies for a fresh state of the
core at the hot zero power level, including the core displacement entirely within the core
basket. The calculations were performed by MCU (Monte-Carlo code) for the full threedimensional reactor core model. It is shown that even small displacements can lead to
significant distortions of the power distribution.

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.
At first the problem is transposed into a linear integral equation for neutron flux. Next a
suitable linear space of functions is chosen to solution of that equation. In this domain,
solution to the integral equation is found by method of successive iterations and its
uniqueness is proven.
Numeric solution to the integral equation by Monte Carlo method consists in finding of one
or several functionals of the exact solution. To this purpose several types of random process
and random variables are constructed in the paper and consequently it is proven there that
these variables are unbiased estimators of those functionals.
Big space of the paper concerns construction of special random variable that can be treated
as a generalization of track length estimator.

The few-group cross section libraries, used by reactor dynamics codes, are affected by
the spectral history effect ? a dependence of fuel cross sections not only on burnup, but
also on local spectral conditions during burnup. A cross section correction method based
on Pu-239 concentration was implemented in the reactor dynamic code DYN3D.
This paper describes the influence of a cross section correction on full-core calculation
results. Steady-state and burnup characteristics of a PWR equilibrium cycle, calculated by
DYN3D with and without cross section corrections, are compared. A study has shown a
significant influence of spectral history on axial power and burnup distributions as well as
on calculated cycle length. An impact of the correction on transient calculations is studied
for a control rod ejection example.

Results of assessment of burnup history effect by moderator density on neutron physical
characteristics of WWER-1000 core are presented on example of stationary fuel loading with
Russian design FA TWSA and AER benchmark for Khmelnitsky NPP that was proposed by
TUV and SSTC NRC at 19th symposium. Assessment was performed by DYN3D code and
cross section library sets generated by HELIOS code.
Burnup history was taken into account by preparing of numerous cross section sets with
different isotopic composition each of which was obtained by burning under different
moderator density. For analysis of history effect 20 cross section sets were prepared for each
FA corresponded to each of 20 axial layers of reactor core model for DYN3D code. Four fuel
cycles were modeled both for stationary fuel loading with TWSA and AER benchmark for
Khmelnitsky NPP to obtain steady value of error due to neglect of burnup history effect.
Main attention of study was paid to effect of burnup history by moderator density to
axial power distribution. Results of study for AER benchmark were compared with
experimental values of axial power distribution for FAs of 1st, 2nd, 3rd and 4th year operation.
The work was performed in framework of orders BMU SR 2511 and BMU R0801504
(SR2611). The report describes the opinion and view of the contractor ? SSTC N&RS – and
does not necessarily represent the opinion of the ordering party – BMU-BfS/GRS and T?V
S?D.

Code package PERMAK-3D/SC-1 was developed for performing pin-by-pin coupled
neutronic and thermal hydraulic calculation of the core fragment of seven fuel assemblies and
was designed on the basis of 3D multigroup pin-by-pin code PERMAK-3D and 3D
(subchannel) thermal hydraulic code SC-1
)
The code package predicts axial and radial pin-by-pin power distribution and coolant
parameters in simulated region (enthalpies, velocities, void fractions, boiling and DNBR
margins). The report describes some new steps in code package development.
Some PERMAK-3D/SC-1 outcomes of VVER calculations are presented in the report.

21st Symposium of AER on
VVER Reactor Physics and Reactor Safety
Dresden, September 19 – 23, 2011
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21st Symposium of AER on
VVER Reactor Physics and Reactor Safety
Dresden, September 19 – 23, 2011
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The ?Full-Core? benchmark ¹ is a 2D calculation benchmark based on the VVER-440
reactor core cold state geometry with taking into the 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 that are placed mainly at the border of the VVER-440 core. Because value of FdH
is not directly measured by the core monitoring system, a decision about a proposal of
benchmark of this kind has been made. In this contribution we define the ?Full-Core?
benchmark and we present the preliminary reference Monte Carlo calculation results.

General review of experimental and calculation researches on VVER-440 and VVER-1000
mock-ups on the reactor LR-0 was introduced on the 20th Symposium AER ¹.
The experimental core fission rate distribution was obtained by means of gamma-scanning
of the fuel pins – 140La single peak (1596 keV) measurements and wide energy range
(approximately 600-900 keV) measurements. Altogether from 260 to 500 fuel pins were
scanned in different experiments. The measurements were arranged in the middle of the fuel
(the active part of pin).
Pin-to-pin calculations of the VVER-1000 mock-up core fission rate distribution were
performed with several codes: Monte Carlo codes MCU-REA/2 and MCNPX with different
nuclear data libraries, diffusion code RADAR (63 energy groups library) and code SVL based on
Surface Harmonics Method (69 energy groups). Calculated data are compared with experimental
ones.
The obtained results allow developing the benchmark for core calculations methodologies,
evaluating and validating source reliability for the out-of-core (inside and outside pressure
vessel) neutron transport calculations.
Key words: SVL, surface harmonics method, VVER, LR-0

The experiments performed at the ZR-6 zero critical reactor by the Temporary International
Collective (TIC) and a numerical assembly burnup benchmark specified for depletion
calculation of a VVER-440 assembly containing Gd burnable poison were used to qualify the
APOLLO2 (APOLLO2.8-E3) code as a part of its ongoing validation activity. The work is
part of the NURISP project, where KFKI AEKI undertook to develop and qualify some
calculation schemes for hexagonal problems. Concerning the ZR-6 measurements, single cell,
macro cell and 2D calculations of selected regular and perturbed experiments are being used
for the validation. In the 2D cases the radial leakage is also taken into account in the
calculations together with the axial leakage represented by the measured axial buckling.
Criticality parameter and reaction rate comparisons are presented. Although various sets of
the experiments have been selected for the validation, good agreement of the measured and
calculated parameters could be found by using the different options offered by APOLLO2. An
additional mathematical benchmark – presented in the paper – also attests for the reliability of
APOLLO2. All the test results prove the reliability of APOLLO2 for VVER core calculations.
1. INTRODUCTION
The main goal of the NUclear Reactor Integrated Simulation Project (NURISP) launched by
the European Union is to make new and significant steps towards a European Reference
Simulation Platform for applications relevant to present power plants including VVERs and
also to future reactors. The individual programs integrated into the platform are to be verified,
validated and demonstrated through several benchmarks corresponding to present and future
PWRs, VVERs and BWRs.
In the frame of this cooperation, MTA KFKI AEKI is performing calculations by using the
APOLLO2 code [1-2]. As a first step, calculations for ZR-6 experiments have been performed
on the basis of the NURISP benchmark specifications and the results have been compared to
the measured values and to the results from the lattice transport module of the code complex
KARATE-440 ¹⁴. The experimental data used in this paper are compiled in the NEA data
bases ³. Calculations on single pin-cells, macro cells and 2D calculations of full

DYN3D is a three-dimensional nodal di?usion code for steady-state and transient analyses
of Light-Water Reactors (LWRs) with square and hexagonal fuel assembly geometries.
Currently, several versions of the DYN3D code are available including a multi-group
di?usion and a simpli?ed P3 (SP3 ) neutron transport option.
In this work, the multi-group SP3 method based on trigonal-z geometry was developed.
The method is applicable to the analysis of reactor cores with hexagonal fuel assemblies
and allows ?exible mesh re?nement, which is of particular importance for VVER-type
Pressurized Water Reactors (PWRs) as well as for innovative reactor concepts including
block type High-Temperature Reactors (HTRs) and Sodium Fast Reactors (SFRs).
In this paper, the theoretical background for the trigonal SP3 methodology is outlined and
the results of a preliminary veri?cation analysis are presented by means of a simpli?ed
VVER-440 core test example. The accordant cross sections and reference solutions were
produced by the Monte Carlo code SERPENT. The DYN3D results are in good agreement
with the reference solutions. The average deviation in the nodal power distribution is
about 1%.

The ?MIDICORE? VVER-1000 core periphery power distribution benchmark was proposed
by Mr. Mikol?? on the 20th Symposium of AER in Finland in 2010. This MIDICORE
benchmark is a 2D calculation benchmark based on the VVER-1000 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-1000 core. In this paper we present our results (keff, integral fission
power) calculated by MCNP and the KARATE code system in KFKI-AEKI and the
comparison to the preliminary reference Monte Carlo calculation results made by NRI, Rez.

SOLUTIONS FOR THE TASK 1 AND TASK 2 OF THE BENCHMARK FOR CORE
BURNUP CALCULATIONS FOR A VVER-1000 REACTOR
1)

APROS – Advanced PROcess Simulation environment ? is a widely used simulation tool for
nuclear power plant modelling. Originally, the three-dimensional neutronics calculation has
been based on a model using the difference method. The model was originally intended also
for power plant simulator applications, where it fits well because of its speed. For safety
analysis purposes, however, a new model was considered to be important improvement to
conform the increasing accuracy requirements. A sophisticated nodal model used already in
HEXTRAN and TRAB-3D was decided to be an implemented into APROS. The model was
reprogrammed into APROS with some improvements and thus, an extensive validation
program was necessary to prove the correct behaviour of the model. Since the model is based
on already-validated HEXTRAN and TRAB-3D, the validation is mainly based on
comparison of the results by APROS to them. The hexagonal part of the model has already
been validated widely and for the square-lattice part, it is undergoing. In the paper, summary
of the latest phases of the implementation and validation results of the model are presented.

AER WORKING GROUP C ACTIVITY IN 2011

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) and two units of Bohunice
NPP (Slovak Republic) replacing the original Russian VK3 system. By both Czech and
Slovak nuclear regulatory bodies it 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 (Upgrade 2 at EBO Slovakia in 2009, Upgrade 5 at EDU
Czech Republic in 2010) the SCORPIO-VVER is still in focus of Central European nuclear
power plants with the roadmap of modification and implementation up to 2015.
In April 2011 the Upgrade 3 at EBO Slovakia has been contracted to support the changed
reactor technical specification and for support of the new type of fuel planned to load in 2012.
During the summer of 2011 the discussions started with EDU NPP in Czech Republic
regarding to the future development of the SCORPIO-VVER system up to 2015.
Parallel with the support of current installations at NPPs the project of new installations is
ongoing. During the years 2012-2013 the SCORPIO-VVER system will become a part of the
full scale simulator in the Centre for training and education of the reactor operators and
reactor physicist for EBO NPP in Slovak Republic.
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 Team of Scientific and Technical Commissioning Support (TSTCS) shall provide an independent
support for the Mochovce NPP 3&4 Commissioning Department during Mochovce Units 3&4
commissioning. This independent support will be in line with the Mochovce NPP 3&4 Directive ?Nonactive tests and commissioning? and it will be carried out in form of professional and expert works focusing
on supervision of fulfilment of requirements for nuclear safety observance. The TSTCS duty to provide for
such services during NPP commissioning is specified by Slovak Regulatory Body legislation.
The independent TSTCS will supervise:
? fulfilment of requirements for nuclear safety during preparation and implementation of commissioning
tests,
? scientific and technical level of commissioning programmes, and reflection on nuclear safety
requirements in commissioning programmes,
? commissioning process and test results.
Main standpoints of the Team activities for individual unit commissioning stages will be:
? assesment of the selected programs of functional tests in instalations having an impact on nuclear
safety and evaluation of the results of these tests,
? assesment of the programs of physics and power commissioning,
? assesment of the unit preparedness before fuel loading start,
? assesment of the unit preparedness for performing initial criticality and low power commissioning and
power commissioning stages,
? evaluation of the results of physics and power commissioning stages and sub-stages,
? final evaluation of the results from implementing the physics and power commissioning stages.
The paper also presents a short description of the Team scope activities, the Team organisation, and a
procedure for issuing of standpoints to individual unit commissioning stages.

Review of AER Working Group B Meeting in Plze?, Czech Republic is given.
Regular meeting of Core Design Group was organized by ?KODA JS a.s. and held at Hotel
?U Pramen?? in Plze?, Czech Republica on May 10?11, 2011, together with Working Group
A. Presented papers (see List of papers and List of participants) covered topics as follows.

15m CYCLE OPTION FOR NPP PAKS OPERATION
Imre NEMES
Paks NPP Ltd.
Paks, PO. BOX. 71. H-7031
Hungary
nemes@npp.hu

The base design of the Russian fuel assemblies TVSA have been under operation at
Kozloduy NPP WWER-1000 reactors since 2004. The old type fuel assemblies TVS-M were
gradually substituted till 2008. The TVSA assembly distinguishes itself with much stronger construction. As a burnable absorber it has a mixture of uranium and uniformly distributed Gd in 6 or
more fuel rods. This enables to increase the safety and effectiveness of fuel cycles.
The experience gained during TVSA fuel implementation on units 5 and 6 and KASKAD
code package validation was presented at the 8th International Conference on WWER Fuel Performance, Modelling and Experimental Support in 2009.
Additional information about TVSA fuel implementation at Kozloduy NPP WWER-1000
units in a 4-year fuel cycle with 42 and 48 fresh fuel assemblies reloading scheme is presented in
the paper.
1. INTRODUCTION
At the end of 2008 the transition process from the old Russian designed TVS-M fuel assemblies to the advanced TVSA fuel type was accomplished at Kozloduy NPP. The old fuel type
had to be changed because of a distortion tendency observed several years ago. The TVSA assembly distinguishes itself with much stronger construction. As a burnable absorber it has a mixture of
uranium and uniformly distributed Gd in 6 or more fuel rods. This enables to increase the safety and
effectiveness of fuel cycles. In the last few years the operational strategy for WWER-1000 units is
about 320 days in full power operation and ~ 45 days of annual outage.
The experience gained during TVSA fuel implementation at Kozloduy NPP units 5 and 6
was presented at the 8th International Conference on WWER Fuel Performance, Modelling and Experimental Support in 2009. The paper presented there mainly paid attention on the KASKAD code
package validation process against the relevant experimental data ¹.
Additional information about TVSA fuel implementation at Kozloduy NPP WWER-1000
units in a 4-year fuel cycle with 42 and 48 fresh fuel assemblies reloading scheme is presented in
the paper.
2. TVSA FUEL CYCLES AT KOZLODUY NPP
2.1. Codes used
The modified Russian code package KASKAD ² mainly consists of the 3D nodal diffusion code BIPR-7A and the 2D pin-by-pin diffusion code PERMAK-A. It has been used for core
design and neutron-physics calculations of WWER-1000 reactors since 2009 at Kozloduy NPP. Its

21st Symposium of AER on
VVER Reactor Physics and Reactor Safety
Dresden, September 19 – 23, 2011
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NEW PRACTICE FOR THE EVALUATION OF ROD EFFICIENCY
MEASUREMENT BY ROD DROP
AT THE NPP PAKS
Istv?n P?s, Tam?s Park?
Paks NPP Ltd
E-mail:pos@npp.hu

For VVER reactor, especially for Dukovany NPP, the optimizing code ATHENA has been
developed in collaboration of SKODA Nuclear Engineering and the University of West
Bohemia for several years. The ATHENA code serves for reducing number of fresh fuel
assemblies, which are necessary to insert into the core for each cycle, and for optimizing the
positions of fuel assemblies. The operation of the ATHENA code for practical use in
Dukovany NPP is at the beginning, but already now some studies by this code are performed.
Ten cycles with new fuel study is one of them. The new fuel has an average enrichment of
4.38% of 235U and is without central hole. Together with inserting new type of fuel the reactor
thermal power was increased from 1444 MWt to 1471 MWt. The results show good
improvement of the length of the cycle including increasing of thermal power.
Besides the fuel with an average enrichment of 4.38%, other similar fuel with average
enrichment of 4.76% is considered. This highly enriched fuel is still only experimental.
Several cycles of this fuel also show better performance, especially the length of the cycle.
The results of the studies are summarized and will be shown below.

21st Symposium of AER on
VVER Reactor Physics and Reactor Safety
Dresden, September 19 – 23, 2011
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To fulfil the task of core linear power monitoring, software Hortica-M, as a part of
ICIS-M, only uses SPND signals and neutron-physics calculation to perform a construed
calculation of power of unmeasurable FAs. Confidence level of interpreting SPND signals to
great extent defines safe and efficient reactor operation.
Up to the present time, the share of reactor gamma radiation in full SPND signal was
assumed not to exceed one percent and not to depend on fuel enrichment and FA or SPND
burnup. Due to the difficulty of experimental estimation of gamma-radiation contribution to
SPND current, which exists nowadays, the paper gives a calculated estimation made with use
of up-to-date computer codes and constant libraries.
On the grounds of experimental data, the paper makes a conclusion about the
expedience of accounting reactor gamma radiation and dependence both on FA type and
burnup in FA and SPND in the conversion function from SPND current to the power density
of six rods surrounding SPND.

The AER Working Group D on VVER reactor safety analysis held its 20th meeting in
Stockholm, Sweden, during the period 12-13 April, 2011. The meeting was hosted by the
Royal Institute of Technology (KTH) and was held in conjunction with the third workshop on
the OECD/NEA Benchmark for the Kalinin-3 VVER-1000 NPP and the fifth workshop on the
OECD Benchmark for Uncertainty Analysis in Best-Estimate Modelling (UAM) for Design,
Operation and Safety Analysis of LWRs. Altogether 18 participants attended the meeting of
the working group D, 12 from AER member organizations and 6 guests from non-member
organization. The co-ordinator 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 gives a brief survey of the 7th three-dimensional AER dynamic benchmark
calculation results received with the codes DYN3D and RELAP5-3D at NRI e?. This
benchmark was defined at the 20th AER Symposium in Hanassari (Finland). It is focused on
investigation of transient behaviour in a VVER-440 nuclear power plant. Its initiating event is
opening of the main isolation valve and re-connection of the loop with its main circulation
pump in operation. The VVER-440 plant is at the end of the first fuel cycle and in hot full
power conditions. Stationary and burnup calculations were performed with the code DYN3D.
Transient calculation was made with the system code RELAP5-3D . The two-group
homogenized cross sections library HELGD05 created by HELIOS code was used for the
generation of reactor core neutronic parameters. The detailed six loops model of NPP
Dukovany was adopted for the 7th AER dynamic benchmark purposes. The RELAP5-3D full
core neutronic model was coupled with 49 core thermal-hydraulic channels and 8 reflector
channels connected with the three-dimensional model of the reactor vessel. The detailed
nodalization of reactor downcomer, lower and upper plenum was used. Mixing in lower and
upper plenum was simulated. The first part of paper contains a brief characteristic of
RELAP5-3D system code and a short description of NPP input deck and reactor core model.
The second part shows the time dependencies of important global and local parameters.

The 7th dynamic benchmark is the most recent problem specified within the AER activities on
verification of the coupled neutronic/thermal-hydraulic codes. This is the VVER-440 transient
initiated by re-connection of the isolated reactor circulation loop filled with a low-temperature
coolant. During the transient a core power excursion with a non-symmetric radial distribution is
expected due to incomplete coolant mixing at the core inlet.
The externally coupled code system DYN3D/ATHLET developed in HZDR was used in the
benchmark simulation. In the case of external coupling all neutronic and thermal-hydraulic
parameters of the core are calculated by the 3D code of neutron kinetics DYN3D.
The computer model of the VVER-440 reactor used in the AER-7 calculations represents a
modified version of the earlier developed input deck for the AER-6 benchmark simulation. The
model includes 6 reactor loops with the main circulation pumps and steam generators. The
secondary system is modelled up to the turbine valves. The 3D core model includes all 349 fuel
assemblies (assembly-by-assembly approach).
The main results of the benchmark calculation are presented and discussed.

21st Symposium of AER on
VVER Reactor Physics and Reactor Safety
Dresden, September 19 – 23, 2011
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In this paper the approach to the question about the spatial distribution of the parameters of the
coolant in-reactor volume. To describe the in-core space is used specially developed preprocessor.
When the work of the preprocessor in the first place, is recreated on the basis of available information
(mostly – the original drawings) with high accuracy three-dimensional description of the structures of
the reactor volume and, secondly, are prepared on this basis blocks input to the nodal system code
improved estimate ATHLET, allows to take into account the hydrodynamic interaction between the
spatial control volumes. As an example the special case of solutions of international standard problem
on the reconstruction of the transition process in the third unit of the Kalinin nuclear power plant, due
to the shutdown of one of the four Main Coolant Pumps (MCP) in operation at the rated capacity (first
download). Model-core area consists of approximately 58 000 control volumes and spatial
relationships. It shows the influence of certain structural units of the core to the distribution of the
mass floe rate of its height. It is detected a strong cross-flow coolant in the area over the baffle.
Moreover, we study the distribution of the coolant temperature at the assembly head of VVER-1000
reactor. It is shown that in the region of the top of the assembly head, where we have installation of
thermocouples, the flow coolant for internal assemblies core is formed by only from guide channel
Reactor control and protected system (CPS) Control rod (CR) flow, or a mixture of the guide channel
flow and flow from the area in front of top grid head assembly (the peripheral assemblies). It is shown
that the magnitude of the flow guide channels affects not only the position of control rods, but also the
presence of a particular type of measuring channels (Self powered neutron detector sensors (SPND) or
Temperature control (TC) sensors) in the cassette.
The given work represents generalization of the first results under the analysis of spatial
distribution of parameters of the heat-assemblies in reactor pressure volume, received with use of
specially developed spatial model ¹ for inside spaces of reactor WWER-1000 for code ATHLET ².
Object of research is the block ?3 Kalinin NPP during development work on nominal power. The
mode with switching-off of the main circulating pump (MCP) on the first loop is considered at work of
a reactor on 98,6 % of power (assemble TVSA, the first loading, 126 effective days). Discussed results
are included into a circle of the problems making the international standard problem, based on the
specified transient which detailed description is resulted in ³.
The essence of construction of spatial model is clear from the summary resulted above. Rather
well the procedure for model construction is described in works [4-6]. It is necessary to note, that in
this paper is considered only hydrodynamic problem for inside spaces of the reactor. To except of the
uncertainties, connected with calculation spatial neutron kinetic, the experimental data on spatial
power distribution, received directly from NPP during work an investigated time interval are used.
Besides these were used as boundary conditions the experimental dada for the coolant temperature at
the reactor input of corresponding loops.
Here it is necessary to specify, that we shall mean temperature of the coolant entrance data in an
installation site of gauges of temperature on an entrance stream of the coolant, i.e. at a reverser of the
coolant in the loop with switch off MCP the entrance temperature in a reactor for this loop there will
be a temperature in a hot leg , and on working loops ? temperature in a cold leg. Thus, the second
contour is excluded from consideration.
Geometrical characteristics of the main circulating loops, collectors and tubing of steam generators
are completely kept. Maintenance of volume of the coolant in the first circuit it is supervised under
actual readings of the gauge of a level in the pressurizer. Experimental dada for pressure an output of a

The paper describes the performed comparisons of measured and simulated local core data
based on the OECD/NEA Benchmark on Kalinin-3 NPP: ‘Switching off of one of the four
operating main circulation pumps at nominal reactor power’. The local measurements of incore self-powered neutron detectors (SPND) in 64 fuel assemblies (FA) on 7 axial levels are
used for the comparisons of the assemblies? axial power distributions and the thermocouples?
readings at 93 FA heads are applied for the FA coolant temperature comparisons.
The analyses are done on the base of benchmark transient calculations performed with the
coupled system code ATHLET/BIPR-VVER. In order to describe more realistically the fluid
mixing phenomena in a reactor pressure vessel (RPV) a new enhanced nodalization scheme is
being developed. It could take into account asymmetric flow behaviour in the RPV structures
like downcomer, reactor core inlet and outlet, control rods? guided tubes, support grids etc.
For this purpose details of the core geometry are modelled. About 58000 control volumes and
junctions are applied. Cross connection are used to describe the interaction between the fluid
objects.
The performed comparisons are of great interest because they show some advantages by
performing coupled code production pseudo-3D analysis of NPPs applying the parallel
thermo-hydraulic channel methodology (or 1D thermo-hydraulic system code modeling).

Results of calculation assessment of positive reactivity insertion rate are presented
under condition of simultaneous effect on reactivity of two provided by WWER project
reactivity operation systems – extract of regulation group and decrease of boron acid
concentration. Showed that maximal positive reactivity insertion rate is achieved at hot zero
power and it?s too less than limit value 0.07?eff/sec defined by normative document ?Nuclear
safety regulation of NPP with pressurized water reactors?. For given reasons concluded that
necessity of application of the extended protection PZ-2 actuation time is absent on interval of
pure condensate transportation for full exclusion of possibility of positive reactivity insertion
simultaneously by two different operation systems.
The work was performed in framework of orders BMU SR 2511 and BMU R0801504
(SR2611). The report describes the opinion and view of the contractor ? SSTC N&RS – and
does not necessarily represent the opinion of the ordering party – BMU-BfS/GRS and T?V
S?D.
INTRODUCTION
According to the Ukrainian regulation ?Rules of Nuclear Safety for Nuclear Power
Plants with Pressurized Water Reactors? ¹ an introduction of positive reactivity
simultaneously by two or more means of impact on reactivity isn?t permitted. Preventive
protection PP-2 upon valve opening on the feed and bleed line is provided for realization of
this requirement on Ukrainian NPPs with WWER reactors. It forbids extraction of regulating
group CR under pure condensate feed is injected in reactor core. The given action really
excludes simultaneous carrying-out by personnel of operations on introduction of positive
reactivity by two or more means of impact on reactivity. However possibility of introduction
of positive reactivity by two means of impact on reactivity after realization of this action
remains, because upon valve closing on a feed and bleed line action of protection PP-2 stops,
but transportation of a pure condensate in reactor core proceeds. Transportation time can
amount to 15 minutes depending on the flow of the feed and bleed line. Therefore during this
time interval the situation is possible when positive reactivity can be introduced

The research field focussing on the investigations and the analyses of severe accidents is an
important part of the nuclear safety. To maintain the safety barriers as long as possible and to
retain the radioactivity within the airtight premises or the containment, to avoid or mitigate the
consequences of such events and to assess the risk, thorough studies are needed. On the one
side, it is the aim of the severe accident research to understand the complex phenomena
during the in- and ex-vessel phase, involving reactor-physics, thermal-hydraulics, physicochemical and mechanical processes. On the other side the investigations strive for effective
severe accident management measures.

A Five-Year Core for a Small Modular Light Water Reactor
Alexey Soldatov and Todd S. Palmer
Oregon State University, Dept. of Nucl. Engr. and Rad. Health Phys., 116 Radiation Center
Corvallis, OR 97331-5902,
soldatoa@onid.orst.edu, palmerts@ne.orst.edu

21st Symposium of AER on
VVER Reactor Physics and Reactor Safety
Dresden, September 19 – 23, 2011
Das gew?hlte Dokument war bei Ende des Symposiums leider nicht verf?gbar.
Bitte fragen sie beim Organisator der Veranstaltung nach.
The required document was not yet available at the end of the symposium.
Please contact the organizing committee.

RUTA-70 is a pool type reactor with the rated power of 70 MW designed as a district heat
supplying facility. The basic design principles of this reactor are simplicity of the design and a
high safety level due to a low pressure and a large coolant inventory in the primary system.
The core design with the CERMET fuel rods also contributes to the reactor safety due to a
high thermal conductivity of the fuel matrix and its role as the additional barrier to the fission
products release.
RUTA-70 model for simulations with the internally coupled code systems DYN3D/ATHLET
and DYN3D/RELAP5 was developed. A 3-D power distribution in the core is calculated by
DYN3D with thermal-hydraulic feedback from the system codes. The reactor facility model
includes reactor pool, two reactor loops, circulation pumps, primary and the secondary heat
exchangers and core channels. The core model consists of 16 fuel assemblies in the 60-degree
symmetry sector with different burnable absorber (Gd) concentrations.
The steady-state with the rated reactor parameters and a BDB accident scenario were
simulated with the DYN3D/ATHLET and DYN3D/RELAP5 coupled code systems to verify
these codes. The simulated accident is initiated by the trip of two operating primary
circulation pumps. Additionally to the initial event the SCRAM failure is postulated.
The compared codes give close predictions for the initial and final states of the accident but
not for the transition between them. The observed deviations are explained by differences in
the subcooled boiling models of the system codes ATHLET and RELAP5. Both simulations
confirm a high level of the reactor inherent safety. The allowed safety margins were not
reached.

AER Working Group E in 2011
Vladim?r Chrap?iak, VUJE, Inc., Slovakia, chrapciak@vuje.sk

The paper presents results of nuclear criticality safety analysis of spent fuel storage for the 1st
and 2nd unit of NPP Mochovce. The spent fuel storage pool (compact and reserve grid) was
modeled using the Monte Carlo code MCNP5. Conservative approach was applied and
calculation of keff values was performed for normal and various postulated emergency
conditions in order to evaluate the final maximal keff values. The requirement of current safety
regulations to ensure 5% subcriticality was met except one especially conservative case.

Reutilisation of fissile nuclides after reprocessing in light water reactor VVER-440 with inert
matrix fuel (IMF) is analysed. Development of transmutation assembly, containing two pin
types – with metal molybdenum + PuO2 mixture and with classical UO2 is described. Static
characteristics of five-year VVER-440 fuel cycle with IMF are given. Results of control rod
ejection accident analysis are presented. Necessary VVER-440 IMF fuel improvements are
mentioned.

This paper reports on an investigation into fuel design choices of a PWR operating in a selfsustainable Th-233U fuel cycle. Achieving such self-sustainable with respect to fissile material
fuel cycle would practically eliminate concerns over nuclear fuel supply hundreds of years into
the future. Moreover, utilization of Light Water Reactor technology and its associated vast
experience would allow faster deployment of such fuel cycle without immediate need for
development of fast reactor technology, which tends to be more complex and costly. In order to
evaluate feasibility of this concept, two types of fuel assembly lattices were considered: square
and hexagonal. The hexagonal lattice may offer some advantages over the square one. For
example, the fertile blanket fuel can be packed more tightly reducing the blanket volume fraction
in the core and potentially allowing to achieve higher core average power density. Furthermore,
hexagonal lattice may allow more uniform leakage of neutrons from fissile to fertile regions and
therefore more uniform neutron captures in Thorium blanket. The calculations were carried out
with Monte-Carlo (MC) based BGCore system, which includes neutronic, fuel depletion and
thermo-hydraulic modules. The results were compared to those obtained from Serpent MC code
and deterministic fuel assembly transport code BOXER.
One of the major design challenges associated with the SB concept is high power peaking due
to the high concentration of fissile material in the seed region. In order to explore feasibility of
the studied designs, the calculations were extended to include 3D fuel assembly analysis with
Thermal-Hydraulic (TH) feedback. The coupled neutronic – TH calculations were performed with
BGCore code system. The analysis showed that both hexagonal and square seed-blanket (SB)
fuel assembly designs have a potential of achieving net breeding. While no major neutronic
advantages were observed for either fuel lattice arrangement.

ALLEGRO – INTRODUCTION TO GFR

21th SYMPOSIUM of AER on VVER Reactor Physics and Reactor Safety
September 19 ? 23, 2011, Dresden, Germany

21th SYMPOSIUM of AER on VVER Reactor Physics and Reactor Safety
September 19 ? 23, 2011, Dresden, Germany

SUMMARY ON THE ACTIVITY OF WORKING GROUP G IN 2011

A detailed and complex RPV model of VVER-440/213 type reactor was developed in
Budapest University of Technology and Economics Institute of Nuclear Techniques (BME
NTI) in the previous years ⁴. This model contains the main structural elements as inlet and
outlet nozzles, guide baffles of hydro-accumulators? coolant, alignment drifts, perforated
plates, brake- and guide tube chamber and simplified core. With the new vessel model a series
of parameter studies were performed considering turbulence models, discretisation schemes,
and modeling methods with ANSYS CFX.
In the course of parameter studies the coolant mixing was investigated in the RPV.
The coolant flow was ?traced? with different scalar concentration at the inlet nozzles and its
distribution was calculated at the core bottom. The simulation results were compared with
Paks NPP measured mixing factors data (available from FLOMIX project ³). Based on the
comparison the SST turbulence model was chosen for the further simulations, which unifies
the advantages of two-equation (k-? and k-?) models ⁴.
The most widely used turbulence models are Reynolds-averaged Navier?Stokes
(RANS) models that are based on time-averaging of the equations. Time-averaging filters out
all turbulent scales from the simulation, and the effect of turbulence on the mean flow is then
re-introduced through appropriate modeling assumptions.
Because of this characteristic of SST turbulence model a decision was made in year
2011 to investigate the coolant mixing with improved turbulence model as well. The hybrid
SAS-SST turbulence model was chosen, which is capable of resolving large scale turbulent
structures without the time and grid-scale resolution restrictions of LES, often allowing the
use of existing grids created for RANS simulations.
As a first step the coolant mixing was investigated in the downcomer only. Eddies are
occurred after the loop connection because of the steep flow direction change. This turbulent,
vertiginous flow was investigated with the SAS-SST turbulence model. The first results of
this research are described in the paper.

In this paper the model of a fuel pin of the Training Reactor of Budapest University of
Technology and Economics was investigated with Particle Image Velocimetry and Laser Induced
Fluorescence measurement methods. An experimental setup was designed, built and optimized to
investigate the natural convection around a model of a fuel pin of the Training Reactor. The
processes were analysed using an electrically heated rod, which models the geometry of the fuel
rods in the Training Reactor. The heated length of the model is the same as the active length of
the real fuel rods. The rod is placed in a glass tank with a shape of a square-based prism. An
additional cooling system ensures constant flow conditions around the rod. The setup consists of
an additional flow channel box, the equivalent diameter of which is equal to the equivalent
diameter of the real fuel assembly. Simultaneous measurements of velocity and temperature
fields were performed in different vertical positions for both cases of natural convection with and
without the flow channel box. The effect of the presence of the channel was analyzed, and a
laminarizating influence was observed. The local heat transfer coefficient was calculated for
every measurement. The two dimensional measurement techniques gave extensive results, the
structure of the hydraulic and thermal boundary layer were fully analyzed.

In the Loviisa 1 and 2 VVER-440 type nuclear power plants, the enthalpy rise margin of the
hottest subchannel is monitored during normal operation. The temperature of the coolant in
the hottest subchannel of the fuel assembly is limited to the saturation temperature. This limit
prevents the coolant from reaching bulk boiling and, in addition, prevents excessive crud
build-up caused by subcooled boiling. Coolant temperature inside the fuel assembly is
affected by mixing of the flow inside the fuel rod bundle.
Striving for higher discharge burn-up via higher enrichment and four batch loading leads to
increased reactivity differences between adjacent fuel assemblies in the core. This results in
lower safety margins in the hottest fuel assemblies. In order to retain higher margins the
mixing needs to be enhanced. This is possible by using mixing vanes on the spacer grid outer
strips.
In the future Fortum is planning to order the fuel from the supplier with mixing type of spacer
grids. Already in September 2011 a dummy assembly (fuel pellets substituted with steel)
including mixing vanes on spacer grids was loaded into the core of Loviisa 1. The purpose
here is to become convinced with the mechanical endurance of the assembly. The first fuel
assemblies with mixing vanes are planned to be loaded into the core during the refueling
outage in 2012.
The effect of the different mixing vane modifications for the second generation TVEL 4.4 %
fuel assembly which includes six fuel rods with burnable poison, gadolinium (Gd), is studied
with computational fluid dynamics (CFD) simulation. Goal of the modelling is to find vane
modifications with which sufficient mixing is reached with acceptable increase in the spacer
grid pressure loss. The results of the studies are discussed in the paper.

Three dimensional CFD model has been built to investigate the coolant flow in the
central tube of the VVER-440 fuel assemblies. The model was verified based on measured
data of the Kurchatov Institute ⁵. With the model calculations were performed for two fuel
assemblies used in Paks NPP. One of them has symmetrical and another has inclined pin
power profile. Ratios of the outlet mass fluxes of the central tube to the inlet mass fluxes of
the rod bundle were determined. Heat up ratios of the tube and rod bundle flows were
calculated too. Sensitivity of the results on the assembly power distribution, inlet temperature
and mass flow rate was investigated. The results of these simulations can be used as boundary
conditions of central tube in studies of coolant mixing in fuel assembly heads.

ATHLET simulation results are compared with the available experimental data of C1
experiment on UPTF facility. Leading phenomena by C1 test are the fluid-fluid mixing and
condensation in the cold legs and in the downcomer of the UPTF. The initiating event is the
actuation of the emergency core cooling (ECC) system. Water is injected into the water-filled
or into steam-filled cold legs. The flow mixing and condensation phenomena determine the
global cooldown and the fluid temperature distribution in the downcomer and in that way can
be defined the boundary conditions for performing Pressurized Thermal Shock (PTS)
analysis.
Developed is a detailed ATHLET model of the UPTF based on a special pre-processor which
generates automatically pseudo-3D nodalization scheme of the facility and also the
corresponding ATHLET input data set including a description of the thermo-hydraulic
objects and the connections between them. In order to account for the specific structure of the
facility newly developed advanced capabilities of ATHLET system code are applied.
Moreover in the special version of ATHLET 2.2B several new enhancements are introduced
which give a substantial acceleration of the calculations for a large number of control
volumes and makes feasible transient simulations with a detailed (ten thousands of control
volumes) description of the investigated facility. The presented study is a part of the
validation of the new ATHLET developments and the new nodalization methodology.

A PROPOSAL TO CONTINUATION OF VVER-440 FUEL ASSEMBLY
HEAD BENCHMARK

The paper reports on Moscow meeting of working group H, which was held in
?Kurchatov Institute?, June 28-29. The meeting was attended by 26 physicists from 13
institutions of Hungary, Czechia, Slovakia, Germany, Russia and the Ukraine.
Name-list of the participants and titles of their presentations may be found in the paper, which
also gives brief information on the designation of presentations.

The paper informs the participants of the fact that working group H elaborated a draft
document ?Estimation procedure for engineering margin factors of VVER fuel cycles?. The
document represents recommendations and, if approved, shall be intended for estimation of
engineering margin factors of VVER-440 and VVER-1000.
To enable certain approaches, the document proposes a number of alternative solutions.
The document will be distributed at the Symposium and may be discussed by the members to
issue a final revision.

The paper discusses the feasibility of accounting for uncertainties of control measurements in
estimation of design margin factors. The feasibility is also taken into consideration
proceeding from the fact how much the processed measured data were corrected by a priori
calculated data of measurable parameters. The possibility and feasibility of such data
correction is demonstrated by the authors with the help of Bayes theorem famous in
mathematical statistics

The paper discusses the way for accounting the uncertainties of calculated VVER-1000 rod
powers caused by the changes of fuel assembly shape in the course of operation. The trouble
is that the gap affected power distribution does not obey the normal distribution law, while the
dispersion does not influence the total uncertainty in power distribution. The paper proposes
the methods for accounting the uncertainty and gives the exercise.