28th Symposium of AER on VVER Reactor Physics and Reactor Safety

Welcome talk by the organizers and practical information.

Farewell talk from the organisers and practical info on presentation / paper distribution.

An ACMFD formulation of the HEXNEM3 nodal flux expansion method for solving the two-group neutron diffusion equation is developed. In the time-dependent case a modal decomposition through matrix diagonalization in the energy domain is applied in order to avoid iteration on the group sources when a fully implicit scheme in time is required. The HEXNEM3 nodal expansion model with transverse integration is applied to the modes. The ACMFD coupling coefficients are derived from the scalar flux and net current boundary and continuity conditions. The balance equations are solved simultaneously for the entire three-dimensional two-group problem. Either the node-averaged fluxes or the node-averaged modes can be chosen as nodal unknowns. The resulting nonhomogeneous linear algebraic system allows for a free choice of any appropriate stationary or non-stationary solution method. A particular advantage of the node-averaged modes balance option is that the system matrix is closer to diagonally dominant and simple preconditioning techniques can be used for convergence acceleration.

The new types of nuclear fuel introduction on Ukrainian NPPs for improve operational characteristic and diversification of nuclear fuel supplier are actual process.

A significant effort has been made to license alternative fuel designs for the VVER plants in Ukraine. As a technical support organization of the Ukrainian regulatory authority, SSTC NRS takes part in the licensing process for introduction of new FA types. For carrying out of independent confirmatory calculations of core characteristics the DYN3D code are used. Considering this adaptation of the DYN3D code to new types of fuel is a necessary part of state-of-art calculations. One of the DYN3D code models that need to be modernized for new types of fuel is the simplified model of the gas gap behavior at the fuel burnup. This model in code is represented as a function gap=f(bur) and calculated by TRANSURANUS code, which takes into account material and geometric characteristics of fuel pins. At present for Ukrainian NPPs actual the following fuel types (operated now or perspective): TVEL Fuel Company – FA Second Generation (VVER 440), FA with “slim fp” (VVER 440), TVSA-12 (VVER 1000); Westinghouse – TVS-WR (or RWFA, VVER 1000), FA with IFBA fuel (VVER 1000), “ESSANUF” FA (VVER 440).

Calculation result of change of fuel-cladding gap width depending on burnup by TRANSURANUS code are presented in this report.

The work was performed in the framework of the project BMU/GRS 4717R01520 under German BMU support. The report describes the opinion and view of the contractor – SSTC NRS – and does not necessarily represent the opinion of the ordering party – BMU/GRS and TÜV SÜD.

Studsvik has recently extended the CASMO5 advanced lattice physics code for the analysis of VVER 1000 and 1200 reactors. These extensions form the basis of CASMO5-VVER, which is primarily intended to compute homogenized nodal data for SIMULATE5-VVER. CASMO5-VVER leverages the latest nuclear data and numerical methods, developed for Pressurized Water Reactors (PWRs), to VVER analyses. The current CASMO5 data library, based on ENDF/B-VII.1 nuclear data evaluation, features a detailed 586 energy group structure and more than one thousand unique nuclides and materials. Resonance self-shielding, based on the Equivalence Theory and an Optimal Two-Term Rational (OTTR) method, has been extended to support hexagonal geometry. The solution to the two-dimensional transport equation over a single lattice, or alternatively multi-assembly domains, is based on the new Linear Source (LS) approximation for the Method of Characteristics (MOC), which supports mirror, periodic, and vacuum boundary conditions. The acceleration of the LS MOC solution is attained through the implementation of a Coarse-Mesh Nonlinear Diffusion (CMND) acceleration. Explicit tracking of actinides and fission products is made possible by a detailed description of burnup chains and the implementation of the Chebyshev Rational Approximation Method (CRAM), used for the solution to the Bateman system of equations. Results from preliminary validation against various critical experiments are presented in this work.

Neutronic benchmark based on previous activities with ALLEGRO – gas cooled fast reactor demonstrator – was defined in the framework of EC project VINCO. This assembly based benchmark was solved by HELIOS and SERPENT codes and results were compared with other solutions. Main comparison results and conclusions are included in the paper as well.

Nowadays, due to the growth of computing power and the possibility of parallel computations, there is a tendency to more and more accurate and complete modeling of physical processes. However, many nuclear reactor design calculation programs and models embedded in them do not take into account the exchange between the elementary cells of the reactor. This can greatly affect the results of calculations, especially on the distribution of temperatures and densities of the moderator. In this paper, the results of the coupled thermal-hydraulic and neutron-physical pin-by-pin simulation of the VVER-1000 reactor fuel assembly ¹ obtained using the precision neutron-physical code MCU (Monte-Carlo Universal) ² and best estimate thermal-hydraulic code ATHLET (Analysis of Thermal-hydraulics of Leaks and Transients) ³ with different hydraulic exchange models between elementary cells in this fuel assembly.
This work is a continuation of the previous calculations presented in ⁴. The aim of this work is to determine the importance of taking into account turbulent and convective exchange between the cells of pins in the fuel assembly. For the calculations, high-precision thermo-hydraulic and neutron-physical codes were used with auxiliary programs written in Python 3 and allowing data transfer between input and output code files without changing the structure of the codes themselves.

The development of an entirely new computational framework for coupled core physics problems, called Kraken, has been started at VTT Technical Research Centre of Finland Ltd. The framework consists of modular neutronics, thermal hydraulics and thermal mechanics solvers, and is based on the use of continuous-energy Monte Carlo reactor physics program Serpent. Ants is a new reduced order nodal neutronics program developed as a part of Kraken. The published methodology and first results of Ants has previously been limited to rectangular geometry steady state multigroup diffusion solutions.

This work describes the solution methodology of Ants extended to hexagonal geometry steady state diffusion solutions. The first results using various two-dimensional and three-dimensional hexagonal geometry numerical benchmarks are presented. These benchmarks include the AER-FCM-001 and AER-FCM-101 three-dimensional VVER-440 and VVER-1000 mathematical benchmarks. The obtained effective multiplication factors are within 18 pcm and the RMS relative assembly power differences are within 0.4 % of the reference solutions.

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 sixth meeting was hosted by VÚJE, a.s. in Modra – Harmónia, Slovakia, during the period of April 23 to 25, 2018.

In total 17 participants from 10 member organizations took part in the meeting and 12 papers were presented. Objectives of the meeting of WG A are: Issues connected with spectral calculations and few-group libraries preparation, their accuracy and validation. The presentations were devoted to some aspects of transport and diffusion calculations, libraries preparation, and to the benchmarks dealing with VVER-440 and VVER-1000 core periphery power tilt.

In frame of WG A: Y.P. Kalinin (co-authors S.S. Aleshin and D.S. Oleynik) presented paper “First experience of solving “’Full-Core VVER-440 RK3+ BENCHMARK’ by KASKAD program package”, P. Mikoláš (co-authors J. Vimpel, D. Sprinzl and S.S. Aleshin) spoke about “Full-Core-RK3+ benchmark“, A. Keresztúri (co-authors I. Pataki and B. Batki) discussed “Application of DFs for the VVER-1000 MIDICORE benchmark”, D. Sprinzl (co-author J. Závorka) added presentation “Full-Core VVER-1000 calculation benchmark“, J. Bajgl (co-author M. Bárta) showed “Some Remarks to the Subcriticality Calculation by MOBYDICK Code” and finally P. Dařílek supplied presentation “HELIOS validation at fast spectrum”.

During the meeting there were discussed the questions connected with the benchmarks solutions and, among others, future activities, which are also shortly described at the end of the paper.

The ALLEGRO gas-cooled fast reactor is currently under development. The current core design should be optimized to fulfil several criteria. The optimization from reactor physics point of view requires numerous full core calculations, thus nodal methods should be used. Reliable group constants should be determined to get precise nodal results, with special regard to thermal effect. The whole methodology should be verified.
In this paper, a group constant (GC) generation methodology is shown using the Serpent Monte Carlo code for the ALLEGRO reactor. Thermal effects, such as Doppler and thermal expansion effects were also investigated. These are particularly important since these phenomena are largely different in fast reactors than in thermal systems and play a significant role during transients. The obtained GCs were applied in the KIKO3DMG nodal code. The results of the nodal calculations including reactivity coefficients were verified through the comparison of the full core nodal and Monte Carlo results.
Using the KIKO3DMG code and the determined GCs, the ALLEGRO core was modified to fulfill the following criteria: enough reverse reactivity needed for the determined fuel cycle length; the number of fuel assemblies and the radial peaking factor should be set not to exceed the maximum allowed linear heat rate; displacement per atom quantity should be maximized for the experimental channels. For the above goals, the plutonium content of the fuel assembly was varied. The modified core should be further investigated from thermal hydraulics and reactor physics point of view to get a core, which fulfils all the required criteria.

The C-PORCA/HELIOS models have been used at NPP Paks as basic core neutron physics calculation tools for many years. During the last decade some new kind of fuel assemblies were utilised in Paks. In order to fulfil the accuracy and performance requirements of the off-line core analysis and in-core monitoring continuous development and testing of the codes have been performed. As a latest development the C-PORCA is available to calculate not only VVER-440, but VVER-1000 core as well.

The C-PORCA is a node-wise diffusion model for the purpose of 3D core analysis. As it is common in node-wise approach the assemblies inside the core are divided into hexagonal or triangular prisms with several numbers of axial layers. These space elements called nodes have got homogenous neutron cross sections. The solver of the diffusion equation needs to determine the space dependent neutron flux inside the nodes and join the fluxes between adjacent nodes applying the flux discontinuity and neutron current continuity. For this purpose the hybrid finite element method is used in C-PORCA.
The HELIOS is a well-known neutron transport code. Its utilisation at Paks NPP has dual use. On one hand this code is a basic tool for preparation of homogenised few-group neutron cross sections inside fuel nodes and areas without fuel such as regions of reflectors and absorbers. On the other hand the flexibility of the code allows to use it for the purpose of testing.
In this paper the main characteristics of the diffusion solver applied in the C-PORCA model is described. The accuracy of this solver is also demonstrated, on the basis of comparisons with different international references available in hexagonal geometry. The C-PORCA results have been compared against benchmark data, produced in the frame of the AER (Atomic Energy Research) community during the last decades. All presented comparisons illustrate that the accuracy of the C-PORCA diffusion solver is excellent.

The main objective of the McSAFE project is the development of the Monte Carlo based multi-physics coupled methodologies for reactor analysis and safety investigations of different reactor systems. Key-research areas are e.g. advanced depletion methods, optimal coupling of MC-codes to thermal-hydraulic solvers, time-dependent Monte Carlo and methods and algorithms for massively parallel simulations.
The project has started in September 2017 under the coordination of KIT. Among the othe partners are european research institutes and technical support organisations (VTT, JRC, CEA, HZDR, NRI, KTH, WOOD), electricity providers (CEZ, Preussen Elektra) and consultants (DNC).
The software developped within the project should allow for the high precision evaluation of core safety parameters and will be applicable also to VVER reactor types.

UANDREA is an integrated framework for evaluation of uncertainties in neutron-physical characteristics of VVER reactors. It consists of a tool for nuclear data and material/geometry uncertainties propagation into homogenized cross-sections (using total Monte Carlo approach), wrapper for the core physics code ANDREA to run a series of calculations and finally an automated evaluation tool which computes the respective statistics along with the covariance matrices as well.

The results can be furthermore visualized in the new VIZAR tool which is shipped with the ANDREA code.

The methodology, implementation and some of the results (based on evaluation of several consecutive cycles of VVER-1000 Temelin NPP) are presented.

New code designed to simplify the reference core calculations using the SERPENT
Monte-Carlo code named SFullcore (Serpent fullcore automated processing) is being developed in NRI.
The code is designed for automated management of SERPENT fullcore calculations for VVER 440 and VVER 1000 reactors for one single state of the reactor. It is possible to evaluate states corresponding both to zero power and to on-power operation. The isotopic composition of the fuel with non-zero burnup is provided by pre-calculated libraries.

Good agreement is achieved between the SFullcore predictions and the experimental data.

Studsvik’s in-core fuel management system CMS5-VVER contains lattice physics code CASMO5-VVER and 3D, steady-state nodal code for the hexagonal geometry, SIMULATE5-VVER. The nodal solver of SIMULATE5 along with preliminary benchmarking activities were presented in last year’s AER meeting.

The first part of this paper outlines the calculation of pin-powers in SIMULATE5-VVER: Heterogeneous pin powers are calculated by modulating multi-group homogenous pin powers obtained from the nodal solver of SIMULATE5-VVER with pin form factors, typically in two-groups, from single-assembly CASMO5-VVER lattice calculations.

The second part of the paper demonstrates the accuracy of the nodal and pin power solutions obtained with CMS5-VVER for two numerical benchmarks that are well known in the AER community: the FULLCORE and MIDICORE VVER-1000 benchmarks. These benchmarks, developed by the Skoda JS a.s. group, model the realistic initial cores of a VVER-1000 reactor at cold state point in 2D dimensional geometry for which the reference solutions obtained with Monte Carlo codes MCNP6 and SERPENT: The core reactivity, assembly and pin-by-pin power distributions computed with CMS5-VVER are compared to the reference solutions.

The implementation of the Low Leakge Loading Pattern (L3P) in Dukovany NPP (VVER-440) led to high differences between expected undercriticality during criticality achievement process and measured one. There was formulated a new approach to the undercriticality inspection during criticality achievement process in previous paper in 2014. This method has been used during start-up in Dukovany NPP since 2012.

Non-uniform distributions of spontaneous fission sources in different loading patterns are described in this paper. The influence of this non-uniformity to core power and ionisation chamber (IC) signal behaviour during criticality achievement process is demonstrated. The improvement of the calculation of start-up parameters and the information about the comparison of calculated and measured parameters are described. Results obtained from both criticality reaching methods (boron dilution, control rods withdrawal) are discussed for different situations during fuel cycle.

The HELHEX code package for WWER-1000 reactor core steady state neutron physics calculations has been developed at the University of Sofia in 2013. HELHEX consists of a 3D two-group nodal diffusion code HEX3DA and a 3D two-group pin-by-pin diffusion code HEX3DP. HEX3DA employs a newly developed nodal method HEXNEM3 [Christoskov, Petkov, 2012] which is an extension to the HEXNEM2 method [Grundmann, Hollstein, 1999] implemented in DYN3D code. The method HEXNEM3 is based on transverse integration and a specific two-dimensional expansion of the intranodal fluxes in the hexagonal plane. The nodal and pin-by-pin XS-libraries are generated using the Helios-1.5 lattice code. The albedo coefficients for the radial and axial boundaries are calculated with the Helios-Mariko system. A visualization module HEX3VI coupled with the thermo-hydraulic module COBRA-4I is also developed as a part of the code package.
The experience of HELHEX code packages validation and TVSA and TVSA-12 fuel implementation at Kozloduy NPP units 5 and 6 were presented at the 8th, 9th, 10th, 11th and 12th International Conference on WWER Fuel Performance, Modelling and Experimental Support in 2009, 2011, 2013, 2015 and 2017. Neutron-physics characteristics for Unit 6, cycle 23rd calculated by HELHEX and KASKAD have been compared with relevant measured data.
This paper contains information about start up physics tests program for Unit 6, cycle 23rd. The base construction of TVSA fuel assemblies has been in operation at Kozloduy NPP VVER-1000 reactors since 2004. The improved TVSA-12 assembly distinguishes itself with increased uranium content, enrichment up to 4.7wt% and increased skeleton stiffness. First loading was done at 2016.

This work represents benchmark calculation which is designed to provide the framework to assess the ability of the traditional multi-physics codes to predict the transient response of the NPP. This benchmark is based on the measurements results obtained while power increasing during Rostov NPP unit 2 commissioning.

During the experiment, the make-up system is switched to a pure condensate mode. The above action leads to a decrease of boron acid concentration in the primary circuit coolant and to subsequent increase of the core power. To keep the power constant the control rods group number 10 is inserted with the small step increment.

Calculation of fuel burnup up to the date of the experiment was made with neutron-physical module Desna of Rainbow-TPP code according to the average daily load schedule provided in the input data. The modeling of the experiment was carried out using the full-scale model of VVER-1000 based on Rainbow-NPP code complex. Two-group neutron-physical cross-section library has been calculated by Sapphire-95 code. Obtained results were compared with the results presented in the benchmark specification.

Geometry models for Monte Carlo transport codes have been using standard constructive solid geometry (CSG). The standard approach is using analytical equations for defining surfaces from which spatial cells are constructed. Both union and intersection operators are available, therefore, arbitrary 3-D geometry can be modeled with CSG. However, this approach can be quite time consuming and possibly error prone for complex models. Monte Carlo transport codes are continuously developed, one of the paths is using CAD-based mesh geometry. MCNP6 features unstructured meshes (UM) created with Abaqus/CAE as geometry description. Attila4MC package for creation of UM geometry from CAD model can be used for MCNP6 models.

VVER-1000 fuel assembly model in UM geometry was created for TVSA-T.mod.2 fuel type. This TVSA fuel type is exclusively operated at Temelin NPP in Czechia. Creation of the model consists of deleting small assembly parts that can be neglected for transport calculations, choosing the size of tetrahedron mesh cells and verification of the model. Basic validation of the model was performed, initially for criticality calculations. In the future, the model will be used for criticality safety analyses, preparation of boundary conditions for diffusion codes and radiation shielding analyzes of spent fuel transport and storage facilities.

This article presents a newly developed coupling between the ATHLET and SKETCH-N computer codes. Firstly the SKETCH-N neutron diffusion code and the ATHLET thermal-hydraulic system code are described. After that the coupling scheme and the coupling program is introduced. The coupling was tested on a VVER-1000 fuel pin cell model. This calculation showed that the coupling is functional. Then the coupling program was applied to calculate the steady-state condition described in the Kalinin-3 VVER-1000 benchmark. The SKETCH-N – ATHLET results were compared to experimental data provided in the benchmark description and to results calculated by SKETCH-N – SKAZKA. SKAZKA is a core thermal-hydraulic module which is coupled with SKETCH-N and was previously verified for VVER-1000 core calculations. The newly calculated results show good agreement with the experimental data and their accuracy is at least as good as the accuracy of the SKETCH-N – SKAZKA calculations. The comparison was made for 2D assembly-wise power density distribution and for 3D power density distribution in the fuel assemblies.

The release of the ENDF/B-VIIR1-based nuclear data library with version 2.02.00 of the HELIOS2 lattice physics code entailed an extensive verification program. Calculational mod- els were developed and compared to both select computational benchmarks and measured re- sults from critical benchmarks, to test the performance of the library and the transport solver enhancements. Preliminary results for relevant reactor physics parameters – core eigenvalue, relative pin power, reaction rates – will be presented for a variety of critical facility models, demonstrating the performance, flexibility, and power of the HELIOS2 code system.

During the start-up of the Unit 1 of Novovoronezh NPP-2 (VVER-1200) the extensive programme of physical tests at hot zero power was carried out. Several tests were designed in such a way to use asymmetrical flux disturbances for measurement of important parameters of the core and control and protection system. These tests include measurements of control rod worth by rod drop and by its slow movement in a different state; measurements of rod group efficiency; emergency protection worth measurement.

This kind of tests is also interesting for the purpose of verification and validation of three-dimensional dynamic codes. The paper presents and discusses results of test simulation by means of ATHLET/BIPR-VVER code, which is now widely used for the VVER safety analysis. Standard system code was equipped with the models of ex-core ion chambers to simulate its currents and reactivity readings, which were the most important experimental parameters of the tests.

The title of this presentation is “Physical start-up tests calculations for Dukovany and Temelín NPP using MOBY-DICK macrocode“. The presentation further extends topic presented on 27th Symposium of AER in Munich. Three tests are discussed: “SCRAM” test for Temelín NPP and “Weight of 6th group of control assemblies” test and “Thermal reactivity coefficient” test for Dukovany NPP. The first part describes MOBY-DICK macrocode and auxiliary functions used for calculations, as well as methodology of individual tests calculations, while the second part presents results of calculations and comparison with experimental values. “SCRAM” test calculation was performed for Temelín NPP Unit 1 Cycle 13 and Unit 2 Cycle 12, “Weight of 6th group of control assemblies” test calculation was performed for Dukovany NPP Unit 1 Cycle 33 and “Thermal reactivity coefficient” test calculation was performed for Dukovany NPP Unit 1 Cycle 29, Unit 2 Cycle 28, Unit 3 Cycle 27 and Unit 4 Cycle 27.

A nuclear reactor is a prolific source of potentially dangerous nuclear radiation. This radiation is unavoidable, since most of the radiation released originates with the fission process itself. In addition to the energetic neutrons, gamma rays are emitted simultaneously with the fission event. To enable personnel to work in the vicinity of an operating reactor, it is necessary to absorb the nuclear radiation released in a thick shield surrounding the core. Even for regions where human access in not permitted during reactor operation, the shielding may be necessary to limit the activation and possible destruction of construction materials or electronic devices. To fulfill the present needs in the Slovak nuclear industry detailed and precise KENO 3D model of the VVER-440/V213 reactor has been developed for criticality, shielding and detector response calculations. This paper investigates several modelling issues associated with VVER-440 criticality and shielding calculations using the SCALE computational system. The model was partially validated by the criticality calculation of the real operational conditions reached on the 310th effective day in Slovak NPP Jaslovské Bohunice unit 4 during cycle 30. Special attention was given to the methodology applied to the determination of the fuel isotopic vectors modelled in one-sixth symmetry core configuration based on which the decay gamma and spontaneous fission neutron source was determined. In case of the effective dose calculation the CADIS-FW variance reduction technique was utilized to decrease statistical uncertainties to acceptable values. To demonstrate the capabilities of the SCALE system and the developed VVER-440 model, two point detectors were placed just behind the dry shielding and one was placed at the outer side of the door leads into the room A004.

The criticality safety criteria require that the effective multiplication factor of the investigated system is less than the defined limits. Therefore, the key issue in any criticality safety problem is to estimate all possible uncertainties and to predict consequent deviation of the calculation from reality. If the calculated value is not equal to its true value bias occurs. The fundamental assumption is that the computational bias is mostly caused by errors in the cross-section data. In addition, the use of random variables in the calculation introduces a non-random bias in the computed result as well. The American National Standards are utilized to predict and bound the computational bias of criticality calculations. These standards require the validation of the analytical methods and data used in nuclear criticality safety calculations to quantify the computational bias and its uncertainty. This paper presents a method for determining the computation bias and bias uncertainty for SCALE KENO-VI code focusing on VVER-440 technology. For this analysis a 3D core model was developed by B&J NUCLEAR ltd. company. Several calculation steps are used to address bias estimation method including sensitivity analysis, uncertainty analyses and cross-section adjustment method. In addition, the neutronic similarity of VVER-440 core to several hundred critical benchmark experiments is evaluated by the use of three integral indices. The database of the benchmark experiment is based on the selection and processing procedure VALID provided by the Oak Ridge National Laboratory and specified in the IHECSBE. The results of all analyses performed are given and discussed in the paper.

Recently, calculation benchmark ‘Full-Core’ VVER-440 has been introduced in the AER community with positive response ¹. Therefore we have decided to prepare a similar benchmark for VVER-1000. The main task of this benchmark is again to test the pin by pin power distribution in fuel as-semblies that are placed mainly at the VVER-1000 core periphery. As value of FdH is not directly measured by the core monitoring system a proposal of similar benchmark for macro-codes for VVER-1000 may be useful as well compared to ¹. The ‘Full-Core’ benchmark is 2D calculation benchmark again based on the VVER-1000 reactor core cold state geometry with taking into the account the geometry of explicit radial reflector. This benchmark was defined on AER Symposium in 2016 in Helsinki. In this contribution we present the overview of available macro-codes results.
¹ V. Krýsl et. al.: “Full-Core” VVER-440 calculation benchmark. Kerntechnik, Vol. 79, No. 4, August 2014.

This work deals with “Full-Core” VVER-440 with PK-3+ fuel assemblies. This benchmark comes from the original “Full-Core VVER-440 benchmark [1, 2]. Loading pattern for this core is very similar to the first pattern of the Mochovce NPP. This core is loaded with fuel assemblies with enrichment of 1.6 %w U235, 2.4 %w U235 and 4.25 %w U235. Difference from benchmark ¹ is that in this new benchmark is 6 PK-3+ fuel assemblies (in the 60° degree core symmetry). Also next difference between this new benchmark, and the benchmark ² is, that this new benchmark doesn’t have inserted control rods. Preliminary reference solution MCNP will be presented on this Symposium and this solution will be compared to SERPENT 2.1.30 code with different nuclear data library. For this purpose, special cross-section library was created by the program makxsf for the specific temperature of 543.15 K. Also reference solution is compared to MCU code ³.

¹ V. Krýsl, P. Mikoláš, D. Sprinzl, J. Švarný, ŠKODA JS a.s.: “FULL-CORE”-VVER-440 core periphery power distribution benchmark proposal, In: 21st Symposium of AER on VVER Reactor Physics and Reactor Safety, Dresden, September 19-23, 2011.
² V. Krýsl, P. Mikoláš, D. Sprinzl, J. Švarný, ŠKODA JS a.s.: “Full-Core”-VVER-440 benchmark extension, In: 26th Symposium of AER on VVER Reactor Physics and Reactor Safety, Helsinki, Finland, 10-14. October, 2016
³ S.S Aleshin, A.S. Bikeev, D.S. Oleynik, A.I. Scherenko, Kurchatov Institute: Calculation of „full core“ VVER-1000 benchmark. Proposal for some new VVER-440 benchmarks.

This paper is devoted to the presentation the new third-plus-generation fuel modified JA-profiling enrichment with 6 U-Gd fuel rods for VVER-440. The optimization of profiling enrichment in the cross-section of the fuel bundle RK3+ fuel assembly with 6 U-GD fuel rods was done in order to reduce the non-uniformity of rod power. Since 2017 year and in this article we have been showing, that changing the location of the U-Gd fuel rods in accordance with the JA- profiling fuel enrichment makes it possible to reduce the non-uniformity of rod power in VVER-440 fuel cycles. The new enrichment of RK3+ fuel allows to develop perspective 15-th month VVER-440 fuel cycles, satisfying modern requirements and to provide a work at the increased thermal power. The types of enrichment profiling and main characteristics fuel cycles are presented in this article.

The concept of development of the two-component nuclear power industry is approved in Russia. Thermal and fast reactors will work in a single closed nuclear fuel cycle. Issues of raw materials, spent fuel reprocessing, discharge of fission products, discharge of minor actinides and their transmutation will be resolved at the same time.
Short-term strategy includes increasing of capacities on the basis of VVER technology development and adjustment of the technologies necessary for closing nuclear fuel cycle.
The efficiency of a closed nuclear fuel cycle in many respects depends on new approaches to development of the actual nuclear fuel and advancement of fuel cycles VVER.
The main requirements to nuclear fuel of VVER-1000 and VVER-440 are safety, reliabilities, economic efficiency and competitiveness.

The main directions of works are presented in the report to support of the solution of the main task:
1. Guaranteeing of geometrical stability of fuel design.
2. Increasing of the operational life of fuel assembly.
3. Increasing of the operational reliability of fuel assembly.
4. Implementation of safety and economical fuel cycles.

Construction of nuclear fuel for VVER will be improved generally in a part:
- increasing in uranium capacity,
- changes of quantity of spacer grids,
- implementation of new types of mixing grids and antidebris filters,
- fuel elements without fixing in a support grid,
- implementation of new fuel and constructional materials.

New types of nuclear fuel have been developed, implemented and successfully operated on the NPP in recent years:
- for VVER-1000 – TBCA-PLUS and TBC-2M and their modifications;
- for VVER-440 – WA and CFA of the second and third generations.

Technical and economic characteristics of the VVER nuclear fuel correspond to the level of world manufacturers of nuclear fuel for PWR power reactors.

The perspective directions of development are:
- for VVER-1000 – fuel assembly of the 4th generation and fuel assembly without fixing of fuel elements in a support grid;
- for VVER-440 –fuel rod with external diameter of 8,9 mm of second generation and modification of working assembly without shroud of third generation;
- fuel cycles with use of fuel of average enrichment higher than 5%;
- the works on creation of new fuel compositions and improvement of zirconium materials.

Eighteen-month fuel cycles may be an option for further development of VVER-440 fuel cycles. 18-month cycles are understood as those with a core operating for about 520–530 effective days at uprated power. Furthermore, an average burnup of 60–65 MWd/kg is advisable for fuel assemblies to be unloaded from the core.

This paper presents the results of development of VVER-440 18-month fuel cycles based on newly-designed fuel assemblies, namely:
- RK3+ – shroudless fuel assemblies with wide angle bars instead of shrouds and optimized pitches between fuel rods;
- Slim Rod – shrouded fuel assemblies with rods whose external diameter was reduced to 8.9 mm;
- RK3+SR – shroudless fuel assemblies with wide angle bars instead of shrouds and rods whose external diameter was reduced to 8.9 mm.

This paper also discusses issues related to using erbium as burnable poison in VVERs-440 and compares neutronic parameters of steady-state cores with uranium-gadolinium and new uranium-erbium fuel. The conclusion is that erbium is a promising burnable poison to be used in VVERs-440.

Review of AER Working Group B Meeting at Modra – pension “Harmónia”, Slovakia is given.

In accordance with the program of The State Atomic Energy Corporation ROSATOM for reducing the cost of electrical power generation at the stages of operation, design and construction of NPPs with VVER, JSC TVEL has carried out a technical and economic study with the involvement of the National Research Centre “Kurchatov Institute” in the use of nuclear fuel enriched above the current limit of 5 % for VVER-1000/1200. Based on the work done, the following conclusions have been drawn:
1) for manufacturing of pilot fuel assembly based on uranium-erbium fuel enriched above the limit of 5 % it is required insignificant upgrade of equipment for transportation and storage of fresh and spent nuclear fuel;
2) confirmed the possibility of radiochemical processing of nuclear fuel enriched above 5 % using existing facilities and those that are under construction in the Russian Federation for processing spent nuclear fuel of VVER-1000;
3) due to reduction in the number of fresh loaded fuel assemblies in fuel cycles with an increased interval of operation between reloads (more than 18 months):
• increase in the natural uranium consumption for the full operational lifetime of the power unit is not required;
• increase in separative work unit for the full operational lifetime of the power unit is not required;
• the volume of spent nuclear fuel is decreased;
• time of loading/discharge and transportation of fresh and spent nuclear fuel is reduced.
Moreover, the use of fuel enriched above 5 % will allow eliminate the significant disadvantage of non-zirconium claddings for the tolerant fuel that is the decrease in multiplying properties due to neutron capture. Increased fuel enrichment margin will make possible to increase the multiplication factor of the tolerant fuel with non-zirconium claddings.

The VVER-1000 life extension and power uprating are the important areas of research and development activities of JSC “Rosenergoatom”. Recent researches have shown that the reactor baffle is the main limiting factor of the life extension of Balakovo NPP units 3 and 4. The fuel cycle optimization is necessary for decrease of reactor baffle radiation dose and NPP life extension at uprated power.
The paper contains the results of the optimization of 18-month low-leakage core loadings at 104 % and 107 % of rated power for Balakovo NPP units 3 and 4, the results of calculation of the radiation dose distribution on the reactor vessel and the baffle are presented. It is shown that the optimization of core loadings reduces maximum and also flattens peaks of the radiation dose on the baffle and allows extending the NPP lifetime up to 60 years at uprated power.

In order to increase the efficiency of the fuel utilization at Dukovany NPP, the idea of a change in the structure of the FA, whose basic characteristic is shroud removal and its replacement with structure called “Karkaz”, was adopted. Against the original proposal of the Russian fuel supplier, this design is characterized, among other things, by the fact that the “angles” at the corners of the FA are longer and “overlap” the outer row of fuel rods, with the exception of one “central”. This solution was designed to reduce transverse flow between the FAs and thereby “de facto” maintain the FA property as an isolated set. In addition, the structure of the angles is significantly thinner than the original shroud, and this allows the increase of the pitch between the fuel rods up to 12.6 mm (instead of standard 12.3 mm spacing) to reduce parasitic absorption in this component and improve neutron moderation, thereby improving the neutron-physical characteristics of such FA, resulting in more energy being obtained from a given amount of the fissionable material. The fuel pin is assumed to be unchanged with respect to pin in the FA type Gd-2M+. (Note: The control assemblies remain shroud-type, but there is a change in fuel pin with Gd2O3, such that this fuel pin will also have a tablet of 7.8/0.0 mm.)
Optimization of PK3+ FAs type of different average enrichment was performed in order to find such a radial enrichment profile in FA with the aim to minimize the non-uniform energy generation in FA (during burning).

In addition, optimization was argued that a better (lesser) energy equalization could be achieved by using a change in the location of fuel pin with Gd2O3 burnable absorber – from the 2nd row to the 3rd row of rods (pins) from the edge of the fuel assembly on the fuel assembly diagonal.

The aim of this study is to achieve a full quadruplicate cycle, each of 15 months (approx. 450 days) at 1475 MWt nominal power.

Preliminary results indicate that fuel assemblies combination PK-3+ and Gd-2M+ does not show any unusual phenomena from the point of view of reactor physics. On the other hand, due to the loading of fuel assemblies in combination of 60+12 and 66+6 pieces after several cycles, 1 FA with a burnout in the range of 40-45 000 MWd/tU remains. This is particularly problematic for batch design, because Athena optimization algorithm places this FA around the core center where frequent fresh or 2-year assemblies are often in positions 3 and 12. As a result, this FA has a high burnout increase during one campaign. This increase may be as high as the limit for burnout of the fuel pin, which is set on 62000 MWd/tU, is exceeded.

This strategy is based on the B1C33 campaign implemented at the Dukovany NPP, which is designed to be 395 FPD days. Already the first cycle, which is loaded with 60 fresh PK-3+ FAs and 12 Gd-2M++ CAs, the reached length at EOR is 424 FPD, which means stretchout 26 effective days. On average, the transition cycles stretchout length is 23.5 effective days. For steady cycles this average value is 19.2 effective days.

VVER-440 reactors have been utilized in Slovakia since 1978. So far, the vast majority of their core loadings were designed in VUJE institute. This paper presents a description of the methods and procedures, which have been used on this purpose in the last decade. Main attention is focused on the calculating tools for core refuelling scheme optimization.

For the in-core monitoring system (ICIS), the NPP project establishes operational limits and limits for safe operation, as well as conditions for safe operation.

The limits are set depending on the power and operability of the main process equipment, and the conditions – depending on the performance of sensors and ICIS equipment, herewith the regulations indicate the permissible time of failure and actions of the operative personnel.

When setting the limits, the uncertainty of ICIS calculations is taken into account.

The importance of ICIS for NPP designs with VVER-1200 reactors is also due to the necessity of implementing core protection according to the peaking factors.

ICIS is currently installed at 2 units of AES-2006 (Novovoronezh NPP-2 and Leningrad NPP-2). Tests during the units startup confirmed the design requirements.

The primary circuit coolant temperature at a core entrance is one of the key parameters during operation of NPP power units with VVER. These parameters are used to monitor and operate according to the requirements of power units safe operation regulations. Therefore, as much attention as their importance is given for the full-scale testing of ICMS during commissioning and operating. Based on the results of full-scale ICMS tests during commissioning Novovoronezh NPP-2 Unit-1 it is informed about the first practical experience of the coolant temperature monitoring at a core entrance of VVER-1200 reactor using thermocouples in the in-core detectors assemblies of KNIT type.

Data about fulfilled tests at all stages of commissioning are given and the main results of the performed analysis are shown. The reliability problems of information obtained by means of the considered method of monitoring are presented and the made decisions during commissioning are noted.

The in-core instrumentation system (ICIS) of the VVER was designed for a safe operation of a whole power unit and a nuclear fuel in particular.

As for now, the last generation of in-core instrumentation systems designed by the NRC “Kurchatov Institute” have been successfully deployed and has been operating at power units of Russian nuclear power plants (Balakovo NPP, Kalinin NPP, Rostov NPP, Novovoronezh NPP and Leningrad NPP) and foreign nuclear power plants (Kozloduy NPP, Kudankulam NPP, Tianwan NPP and etc.).

During its operation, ICIS stores a large amount of data into historical data archives. ICIS historical data archives are the prime source of the key scientific information for tasks such as:

- NPP operation analysis;
- Design and verification of a new reactor related software.

Today the amount of the available ICIS historical data surpasses more than one hundred of fuel cycles, totally exceeding approximately 25 TB of a highly compressed data.

The NRC “Kurchatov Institute” have developed special framework to provide access to the compressed ICIS historical data without a headache of deploying full local copy of the ICIS software. This framework includes:

- Tools for the direct work with data;
- Administrative tools for the maintenance of data storage;
- User’s API;
- Built-in visualization and analysis tools.

By the means of this framework scientists and engineers of the NRC “Kurchatov Institute” can easily access and work with any available on the server historical data via LAN.

The NRC “Kurchatov Institute” have proposed a bunch of machine learning projects focused on a development and implementation of empirical predictive models for ICIS. Some moderate progress was achieved with help of such large amount of data and modern open source data analysis/machine learning tools within Python eco system (mostly for rapid R&D).

After a preparation period with VERONA upgrade and lead test assembly program, a new fuel type was introduced at MVM Paks NPP Ltd. This 4.7 % average uranium enriched assembly type, together with the former 4.2 % uranium enriched fuels, allowed us to lengthen the operating cycles to 15 month. Both fuel types contain gadolinium burnable poison, in six and three pins respectively. All of the four units have been converted to the C15 cycles, and have been being operated without any problems in the last few years.

In this presentation the test results of core design code HELIOS/C-PORCA, which is the basic model of VERONA, is outlined. C15 cycles were entirely investigated with the comprehensive study of measured and predicted (calculated) values of different reactor states.

At first step, in order to prove the capability of the reactivity calculation of the nodal diffusion model, critical boric acid concentrations of different burnup and start up states were calculated and compared with measured values. During the next step of the verification process local in-core parameters were investigated. Measured neutron flux distributions (SPND signals) and coolant outlet temperatures were examined. SPND and TC signals were predicted as a part of the monitoring system. The results of statistical investigations (average differences and standard deviations) for the applied fuel types are also presented.

Working Group C had a meeting joined with Working Group G in Balatongyörök, Hungary , 24-25 May 2018. At the joint meeting 26 people from 9 AER member organisations of 5 countries – such as Russia, Finland,Czech Republic, Slovakia, Hungary – participated. In the 2 days of the program 21 papers were presented, 10 of this belonged to Working Group C subject.

Consider the initial stage of the accident (first 0.5 sec.) caused by the rupture of one of the cold loops (D= 850 mm) of the primary circuit of the VVER-1000 reactor. Instantaneous rupture (0.0001 s) of the pipeline in the area between the reactor and the main circulation pump (MCP) with double-end expiration into the rupture. The produced pressure waves and their propagation in the equipment of the primary circuit of the installation are shown. The parameters of a typical reactor plant V-320 (VVER-1000) are used for the calculation, particularly, the 3rd unit of the Kalinin NPP. All initial data for the calculation were obtained from the data of the international standard problem Kalinin-3 [1-2]. The calculations were carried out using the computational best estimated code “ATHLET”, developed by the society for reactor safety (Gesellschaft für Anlagen-und Reaktorsicherheit-GRS), Germany ³ and certified in Russia for use in calculations to justify the safety of reactors with water coolant ⁴. The considered emergency is included in the list of different types of reports necessary for VVER safety justification ⁵. The obtained results can be used as boundary conditions for the analysis of equipment strength in this transition process.

REFERENCES:

1. V.A.Tereshonok, V.S. Stepanov, V.V. Ivchenkov, V.A. Pitilimov, S.P.Nikonov, Description of a Transient Caused by the Switching-off of One of the Four Operating MCP at Nominal Reactor Power at NPP Kalinin Unit 3, NEA/OECD, July 2008
2.V. A. Tereshonok, S.P. Nikonov, M. P. Lizorkin, K. Velkov, A. Pautz, K. Ivanov, International Benchmark for Coupled Codes and Uncertainty Analysis in Modelling: Switching-off of one of the four operating main circulation pumps at nominal power at NPP KALININ UNIT, 18th Symposium of AER on VVER Reactor Physics and Reactor Safety, Hungary, Eger, Oct. 6-10,2008
3. H. Austregesilo, H. Deitenbeck, A. Langenfeld, J. Scheuer, P. Schöffel, ATHLET 3.1A, Programmer Manual, March 2017
4. Аттестационный Паспорт Программного Средства №350 от 17.04.2014
5. Requirements to the content of the report on the safety justification of nuclear power Plants with a VVER type reactor (with Change № 1 from 20.12.2005.).

The OECD/NEA Benchmark on NPP Kalinin Unit 3 “Switching-off of one of the four operating main circulation pumps at nominal power” ¹ based on real measurements is at its end phase. The core load consists of 163 assemblies of the type “TBCA”. A new benchmark is starting based on measurements from the NPP Rostov Unit 2 ² “Reactivity compensation with diluted boron by stepwise insertion of control rod cluster into the VVER-1000 core”. The Rostov core also has 163 assemblies but of a new modern type “TBC-2M” which differs in construction in comparison with the “TBCA”.

The goal of the performed study is to determine the thermohydraulic differences between the cores due to the implementation of the new assembly type. This work will help the Benchmarks’ Rostov-2 participants to set up more accurate core thermohydraulic models.

The paper compares results of steady state simulations for the two cores (Kalinin-3 and Rostov-2) with the same assembly wise axial power distributions. Calculations are done with an under-development version of the GRS system code ATHLET, which has sub-channel modeling capabilities. Thus, seven assemblies around a control rod location are calculated at the sub-channel (pin-by-pin) level. To ensure that the observed differences only come from the fuel assembly construction, the same axial power density is used for all cases.

¹ V.A.Tereshonok, S.P.Nikonov, M.P.Lizorkin, K.Velkov, A.Pautz, K.Ivanov, International Benchmark for Coupled Codes and Uncertainty Analysis in Modelling: Switching-off of one of the four operating main circulation pumps at nominal power at NPP KALININ UNIT 3, 18th Symposium of AER on VVER Reactor Physics and Reactor Safety, Hungary, Eger, Oct. 6-10, 2008.
² M. Avramova A. Denisenko, M. Denisova, P. Gordienko, K. Ivanov, S. Nikonov, I. Pasichnyk, B. Shumskiy, R. Sizov, K. Velkov, EGMPEBV Benchmark “Rostov-2”, 27th Symposium of AER on VVER Reactor Physics and Reactor Safety, Germany, Munich, Oct. 17-20, 2017.

Work deals with RIA accidents safety assessment problem for VVER reactor, which is operating in load regulating regimes. Model approbation is given by safety assessment for normalized primary power grid frequency regulating regimes (NFR). Frequency regulating regimes leads to turbine load variations and following continuously changing of main reactor facility parameters by feedbacks and regulators actions. Dynamic reactor conditions must be taking into account in safety assessment calculations and needs in specialized methodical approach. Current work gives two solutions. First of them is best estimate approach with full dynamics accounting. Second approach bases on stationary parameters range extension and aims to conservative overlap of possible parameters perturbations by NFR without dynamical initial condition. Dynamical method approbation performs by proposed mathematical model for VVER safety assessment with NFR. Model includes the full scope of calculation and theoretical analysis during NFR transient processes and contains description, algorithm and calculation examples. Safety analysis examples are some RIA type accidents such as control rod ejection, uncontrollable rod group withdrawal and control rod drop.

The report provides a brief description of the fuel assembly pin-by-pin model developed earlier with the KORSAR/GP code. The nodalization of the developed model consists of parallel 1D hydraulic elements (channels). Each fuel rod is modeled as a individual structural element. Two special calculation elements of KORSAR/GP are used in the pin-by-pin model for coolant mixing simulation. They are cross-junction and turbulent mixing element. The first one calculates the convection of the coolant. The second one simulates the interchange of coolant between two channels through some surface. Such a model consists of large number of elements (nodes), that increases time calculation. So the the pin-by-pin model application for calculations is not always justified and necessary.

The purpose of this paper is application field assessment of FA pin-by-pin model. The paper presents the results of the “Control rod ejection” transient simulation. This accident is interesting by power redistributions in the accident FA. The safety margins are calculated with use of «hot channel» model for NPP safety analyses. Only the fuel rods of the core with the highest power are taking into account. The absence of pin power distribution change for FA during transient is compensated by increased conservatism. In this paper the maximum fuel rod power assessment was obtained using of pin-by-pin model.

The AER Working Group D on VVER reactor safety analysis held its 27th meeting in Rossendorf, Germany, during the period 12-13 June, 2018. The meeting was hosted by Helmholtz-Zentrum Dresden-Rossendorf. Altogether 19 participants from nine AER member organizations attended the meeting of the working group D. The co-ordinator of the working group, Mr. S. Kliem, served as the chairperson of the meeting.

The meeting started with a general information exchange about the recent activities in the participating organizations.

The given 13 presentations and the discussions can be attributed to the following topics:

- Safety analyses methods and results
- Code development and benchmarking
- Severe accident analyses
- Future activities

The Working Group decided to include also in future the severe accident analyses into the scope of the activities.

A list of the participants and a list of the handouts distributed at the meeting are attached to the report. The corresponding PDF-files of the handouts can be obtained from the chairperson.

The paper contains a description of conservative analysis of initiating event associated with
uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320 of Temelín
NPP with TVSA-T fuel assemblies at hot full power and four coolant loops in operation. This
event is included in the group of beyond design basis accidents. The failure of operator
intervention to isolate dilution routes at intervals shorter than 30 minutes is assumed. The aim
of analysis is to determine also the time interval which is necessary for interventions leading to
the deterrence of fuel damage. Since the response of the whole NPP system influences
the course of safety important parameters of the reactor core, the calculations were made by
an externally coupled version of the 3D reactor dynamic code DYN3D and the thermohydraulic
system code ATHLET. The two-group neutronic cross section library of homogenized cross
sections parameterized for different thermal-hydraulic states prepared for TVSA-T fuel
assemblies is used. It is shown that, in addition to exceeding the DNBR limit of more than 90
minutes from the start of the transient, the remaining safety acceptance criteria will not be
violated until the end of the calculation.

This work was funded by the Project between ČEZ, a. s. – Temelín NPP and ÚJV Řež, a. s.
(Contract No. 14SMP258).

The development of severe accidents is possible for a large number of scenarios with the reactor core damage. The severe accidents, taking into account impacts introduced by NPP emergency systems (for instance, water injection into the primary circuit) make it necessary to perform a calculational analysis of criticality for the number of severe accident scenarios. Reaching criticality can drastically change scenarios and consequences of the accidents of this type.
For the typical base states of severe accidents, calculational models of scenarios of accident situations are formed, that take into account the following: chemical composition of molten materials, distribution of coolant parameters, temperature distribution of the core components and the core internals, change in the configuration of the elements of the basic state model developed for normal operation conditions. In preparation of calculational models, the design characteristics of RP with V-320 are used as well as the thermal-hydraulic characteristics obtained during the analysis of various stages of the scenarios of severe accidents.
Impossibility of realization of repeated criticality is demonstrated based on the analysis of the neutron multiplication factor for base states.

The paper studies the potential for a self-sustaining chain nuclear fission reaction during the propagation of a severe accident at Ukrainian nuclear power plants with VVER-1000 reactors. The beginning of the third fuel cycle of the Х2 core loaded with TVSA fuel was modelled. Some models for criticality calculation at different stages of a severe accident in the VVER-1000 pressure vessel were developed and corium multiplication properties were calculated.

The severe accident in the VVER-1000 core was divided into seven major stages: intact reactor core, beginning of cladding damage (swelling), cladding melting and flowing down to the support grid, melting of structural materials, homogenization of the materials at the bottom of the reactor vessel, stratification of the corium at the bottom of the reactor vessel, spread of the corium from the reactor vessel. A compensatory measure such as addition of the boric acid solution to the cooling water was analyzed.

The results obtained in the study show, if fuel rods are maintained intact at the beginning of a severe accident, criticality might appear even if the emergency protection rods is triggered. With further propagation of the accident, the corium will be deeply subcritical if water cannot penetrate into its pores or voids. In the case of the formation of pores or voids in the melt with the ingress of water, a criticality may occur.

The paper deals with the results of the Leningrad NPP-2 start-up loss of power test and its simulation with use of KORSAR/GP code. The test was held at the 1st unit with VVER 1200 reactor on the 29th of March.
During the test the steam generators were refilled with EFW system at first. This system is connected to the SG steam space. Therefore injection of water leads to the secondary pressure decrease. This phenomenon is difficult for simulation due to complicated structure of the SG internals.
Also the test provides unique data for the reactor facility cooldown at natural circulation in the primary circuit. The important data for reactor head and pressurizer cooldown were obtained.
Thus the test data provides the validation of several processes such as:
- emergency feedwater injection into the steam generator;
- steam generator filling with “cold” water from AFWS;
- pressure regulation at filled pressurizer;
- cooldown of the reactor head and the coolant circulation beneath the head.
The results of the test were used for the validation of KORSAR/GP code, unit nodalization and equipment models.

This paper presents an attempt to use multiphase CFD code for the prediction of the Departure from Nucleate Boiling (DNB) type of Critical Heat Flux (CHF). Numerical simulations of DNB in boiling flow in vertical tube were performed with the Ansys CFX 18 code. This code can simulate multiphase flow by solving three balance equations for each phase. Boiling on the heated wall can be simulated with Kurul and Podowski model. A set of mathematical models of physical phenomena in boiling bubbly flow was selected and successfully tested on the DEBORA experiments with subcooled boiling flow. The same set of models was then used in the simulations of DNB cases. Simulated cases were based on data from the 2006 CHF look-up table by D. C. Groeneveld.

It was found out that the local criterion for DNB prediction can be based on the following parameters calculated with the CFD code on the wall: void fraction, wall superheating and wall shear stress. The proposed method of DNB detection was so far tested on 16 MPa pressure level on many cases covering a large range of parameters (mass flux 1000 – 7000 kg/m2/s and exit equilibrium quality -0.4 – 0) and it worked with +-10% accuracy in almost all cases.

Presented work was carried out within the Technology Agency of the Czech Republic (TA ČR) project TH02020360 „Modeling of the critical heat flux with computational fluid dynamics codes“.

In this paper, the optimization of the assignment of spent fuel assemblies into final disposal canisters is considered. This application is of essential importance because the final disposal canisters are expansive and there is a limit for canister-wise total heat load, which must not be exceeded.

The study utilizes mathematical optimization algorithms that have been developed by Timo Ranta is his D.Sc thesis (Tampere University of Technology, 2012). In the used formulation, the target of the optimization is to minimize the maximum canister-wise decay heat load at the time of the encapsulation.

The optimization algorithms were implemented using the Python programming language. The paper demonstrates results and findings of a work, where the algorithms were utilized for analyzing some arbitrary final disposal scenarios for present and expected future spent fuel assemblies of Loviisa NPP.

Paper concludes that, despite a huge amount of degrees of freedom, the algorithms are capable of finding practically a global optimum for the problem. The outcome of the work is a software tool, which can be used for further final disposal analyses according to various needs This is considered very beneficial as the final disposal costs play an important role in overall economics of the fuel utilization.

Studsvik’s approach to spent nuclear fuel analyses combines isotopic concentrations and cross-sections, calculated by the lattice transport codes, with core irradiation history data from reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code SNF to predict spent nuclear fuel characteristics. The procedure for verification and validation of SNF is outlined in this paper which includes verification of the decay data as well as the numerical methods they are applied into. The paper presents the applicability of SNF to analyses of spent nuclear fuel in comparisons to well established and recognized code in the spent fuel analyses field, ORIGEN.

The phase-out of nuclear energy in Germany has triggered the demand for an additional solution to dispose of damaged fuel rods, normally remaining in the fuel pond until the final shutdown of the NPP. In order to establish a disposal concept for damaged fuel rods suitable for the needs of German utilities, GNS has developed a first of its kind solution, the Integrated Quiver System for Damaged Fuel (GNS IQ), which can be loaded in the transport and storage casks of the CASTOR® family.

The GNS IQ features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. It can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific needs of the customer. The quiver is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well.

The presentation gives a general overview of the disposal concept and provides a description of the Integrated Quiver System for Damaged Fuel with the focus on criticality safety assessment.

According to IAEA PRIS data most part of nuclear reactors under operation in the world are thermal reactors which consume U-235 in once-through fuel cycle. Such approach results in ineffective resource utilization and dramatic SNF volume growth. However, sustainable nuclear energy system (NES) should provide NFC closing for all hazardous radionuclides to minimize its life-time within NES and to make risk to be proportional to NES capacity, rather than total energy produced. These two basic principles require enough amount of neutrons for both energy generation and hazardous radionuclides transition to fission products. Therefore, taking into account politic, economic and technological risks and uncertainties, these issues can be solved in terms of two-component NES consisting of both thermal and fast reactors ¹. In this work two methods to estimate neutron balance in NES are discussed. The fist method is based on the analysis of nuclear transformation chain due to radioactive decays and neutron induced reactions. The second one is the most complete one and relies on reaction rates comparison. Neutron balance estimation approach is demonstrated for two-component NES case study.

Reference

1. P.N. Alekseev, V.G. Asmolov, A.Yu. Gagarinskii et al. On a Nuclear Power Strategy of Russia to 2050. Atomic Energy, Vol. 111, No. 4, February, 2012 (Russian Original Vol. 111, No. 4, October, 2011)

19th session of the AER Working Group F – «Nuclear Fuel Cycle Prospective and Sustainability» was held by National Research Centre “Kurchatov Institute” in Moscow, Russia, 23 – 26 of July, 2018. 21 participants from Czech Republic, Hungary, Russian Federation and Slovak Republic introduced and discussed 16 reports. These reports cover a variety of topics, including:

- Nuclear energy system structure for sustainable development;
- Advanced VVER nuclear fuel cycles;
- REMIX-technology for VVER closed nuclear fuel cycle
- Nonelectric application of nuclear power and synergy with non-nuclear power technologies;
- Burnup code verification on the basis of irradiated VVER-440 fuel benchmark;
- VVER spent nuclear fuel management strategy;
- Nuclear fuel cycle simulation.

The next 20th session of the AER Working Group F – «Nuclear Fuel Cycle Prospective and Sustainability» is planned to organize in July 2019, Moscow, Russia.

Review of AER Working Group E Meeting in UJV Řež, Czech Republic is given.

Presentation of the current AER website and the tools developed for facilitation of the Symposium organisation (registrations, papers, schedule, book of abstracts etc.). Inspiring talk for future Symposium organisers a well as for everyone who is looking for tools to make their daily work automated and easier.