29th Symposium of AER on VVER Reactor Physics and Reactor Safety

The RK3+ benchmark was created for the AER community. The purpose is to test calculation models against a reference solution, calculated by MCNP. The benchmark contains a new fuel assembly for VVER 440 reactors, RK3+. The assembly has the same type of fuel pin as Gd-2M+, but uses different pin pitch and radial profiling and it has a thinner shroud, with a one pin wide gap on each side.
In the presentation two solutions are presented for the benchmark. It was solved by both C-PORCA (nodal 3D diffusion code) and HELIOS (2D transport code). The results are compared to the reference solution. The compared parameters are KQ and KK for various assemblies and Kinf. The benchmark was already solved by C-PORCA for AER-AB 2019, the results, presented here were updated since then.

The VVER-1000 Full Core Calculation Benchmark was presented in AER Symposium in 2016. Recently there are several solutions of this benchmark made by several reactor physics codes. In this paper the solution of the benchmark calculated by the KARATE code system will be presented in details and some comparisons to the reference solution will be shown.

Thermocouple located near the coolant outlet of the fuel assembly head may indicate a temperature biased by radiation heating and subsequently affect accuracy of incore measurements. The analysis splits into two parts; coupled neutron-photon Monte Carlo transport to determine radiation heating that serves as the input to heat transfer calculation resulting in temperature bias evaluation.

Various nuclear reactions contribute to radiation heating in the active volume of the thermocouple steel. Monte Carlo transport calculation of radiation heating for average fuel assembly in one of latest Dukovany NPP cycles was performed with MCNP code. Seven heating sources were compared; activated top nozzle steel photons, activated block of protective tubes photons, fission neutrons, secondary photons, fission photons, spent nucler fuel neutrons and SNF photons. It was found that dominant reactions are activated block of protective tubes and secondary photons and the analysis results can be easily extended to all fuel assemblies based on average fuel assembly model results. Heating power was compared to values published previously for Paks and Kola NPPs.

Heat transfer calculation were performed with TEPLO code developed in SKODA JS. The code calculates temperature field by finite element method in 2-D geometry. Radiation heating can be considered as negligible if the axial shift between thermocouple active part and thermocouple socket is below 15 mm.

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.

The KASKAD software package has been developed in the NRC KI for carrying out neutronic calculations of VVER-type reactors and is widely used in design organizations and at operating NPP units with VVERs in Russia and abroad. Its advanced version named KASKAD 2007 is in trial operation in the NRC KI.
The report provides a brief description of the current state of the software package, including information on its structure, basic calculation and service modules, as well as some materials on their validation.

INFORMATION ON AER WG A ON IMPROVEMENT, EXTENSION AND
VALIDATION OF PARAMETRIZED FEW-GROUP LIBRARIES
FOR VVER 440 AND VVER 1000

Pavel Mikoláš
ŠKODA JS a.s., Orlík 266/15, 316 00 Pilsen, Czech Republic

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 ŠKODA JS a.s. in Pilsen, Czech Republic, during the period of April 15 to 17, 2019.
In total 15 participants from 9 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 assembly and pin power distribution.
In the frame of WG A: D. Sprinzl (co-authors V. Krýsl, P. Mikoláš, J. Vimpel and J. Závorka) presented paper “Reference solution of Full-Core VVER-1000 calculation benchmark”, S. Patai-Szabó spoke about “HELIOS Calculations for ‘Full-Core-RK3+’“, M. Horváth (co-authors I. Pós and T. Parkó) discussed “Fullcore VVER 440 RK3+ benchmark with C-PORCA”, Yu. Kalinin (co-author Aleshin S. S.) added presentation “Results of solving ‘Full-core’ benchmarks by PERMAK-A code“, D. Sprinzl (co-authors V. Krýsl, P. Mikoláš and J. Závorka) showed “Solution of Full-Core VVER-440 PK-3+ calculation benchmark by MOBY-DICK” and finally P. Mikoláš supplied presentation “Uncertainties associated with nuclear data when calculating fuel isotope concentrations at burnup”.
During the meeting were also discussed the questions connected with the benchmark solutions and, among others, future activities, which are shortly described at the end of the paper.

This work deals with the updated reference solution of Full-Core VVER-440 PK-3+ benchmark. The solution has been calculated by Serpent code using Monte Carlo method. And results consist of the effective multiplication factor (k-eff), the power distribution on fuel assembly level (Kq) and on the pin by pin level (Kk). The main aim of this calculation is to create a reference solution to test the pin by pin power distribution for different macro-codes. The solution also includes detailed accuracy analysis of the calculation.

Calculations of neutron-physical characteristics of “Full-core” benchmark by programs TDMCC and Sapfir&RC

Frolova M.V., Antonov S.N., Ustinov A.N., Fateev M.V.

OKB “GIDROPRESS”

In the research the results of verification of program complex Sapfir_95&RC_VVER for calculation of neutron-physical VVER core characteristics was represented.
“Full-core” benchmark was used for verification. Benchmark input data was based on the geometrical and material parameters of Temelin NPP. Support of accuracy of program complex Sapfir_95&RC_VVER calculations was based on using of calculation result, performed by Monte-Carlo method by program TDMCC. Preliminarily calculation results, performed by program TDMCC, were compared with result which were gotten by precision programs Serpent, MCNP and MCU.
In the verification emphasis was placed on the support of accuracy of multiplication factor, FA power distribution in the core and pin-by-pin energy field for all using FA type.
Calculation results showed that program TDMCC can be used as reference, that was confirmed by comparison with programs Serpent, MCNP and MCU. The minimum deviation of calculation results by program complex Sapfir_95&RC_VVER and program TDMCC was gotten for calculation by neutron flux microdistribution superposition method.

The system thermal hydraulics neutronics computer Code “ATMIKA-LWR” is developed with the objective to perform indigenous safety analysis of VVER 1000 MWe reactors of Kudankulam Nuclear Power Plant (KKNPP) of India. The code is developed for analysis of various Postulated Initiating events (PIEs) which include Anticipated operational Occurrences (AOO), Design Basis Accident (DBA) and for Design Extension- A (DEC–A) events of VVER 1000 Reactors.

The code simulates the overall thermal hydraulics of the plant along with integrated control systems viz. Primary pressure control, Steam Generator (SG) pressure and level controls, Turbine Governor controls, Reactor control system along with Emergency protection & Preventive Protection, Safety Injection and other systems & automatic actuation logics. Presently the computer code is coupled with point kinetics model and work is going on for its coupling with 3D kinetics for space effects.

The various simulation studies performed and results of AOOs such as RCP trip, Turbine trip, Reactor trip, Load rejection, TDFP trips etc. are compared with KKNPP Commissioning Test results. The unplanned plant transient events such as 2 TDFP trip with failure of starting of 1 EDFP causing emergency protection trip of KKNPP are simulated using the computer code. The simulation of SG tube rupture event with primary and secondary cooling by systems such as SGECD and EBIS is completed and results are compared with published results of VVER 1000 reactors. This paper presents the modelling aspects and results of simulation studies of 2 Adjacent RCP Trip using the computer code.

Studsvik’s in-core fuel management code package CMS5-VVER, which includes the CASMO5-VVER advanced lattice physics code and SIMULATE5-VVER three-dimensional nodal code, is currently in use for VVER-1000/1200 reactor analysis. Preliminary validation of CASMO5-VVER has been presented for various critical experiments using the ENDF/B-VII.1-based nuclear data library. Analogously, benchmarking of SIMULATE5-VVER for the X2 VVER-1000 benchmark has also been presented in previous publications using the same data library.
Recently, a new commercially available CASMO5-VVER nuclear data library has been generated based on the ENDF/B-VIII.0 evaluation. The ENDF/B-VIII.0 evaluation represents the state-of-the-art in nuclear data and features new incident-neutron cross section evaluations from the CIELO project for 1H, 16O, 56Fe, 235U, 238U and 239Pu. A summary of the main features and validation of the new CASMO5-VVER ENDF/B-VIII.0-based data library, referred to as E8R0 library, are presented in this work. Comparisons of predicted criticality and fission rate distributions to measurements from various hexagonal-lattice critical experiments, such as the ZR-6 (TIC) and P-Facility, along with CMS5-VVER benchmarking of the E8R0 library to the X2 full-core benchmark for Cycles 1 through 4, will be presented and discussed in this work.

The application of the advanced measurement method of CPS CR group worth was studied in the VVER unit of the Tianwan Nuclear Power Station (TNPS), that is, the current value of the detector outside the reactor was recorded during fast insertion and lifting of a certain group of control rods (the core boric acid concentration is unchanged). After considering the correction of the spatial correction factors (dynamic spatial factor and static spatial factor), the integral value of each group of control rods can be accurately measured, as well as the value of emergency protection, and the deviation between the measurement results and the theoretical calculation values meets the requirements. The four VVER-1000 units in operation at TNPS have successfully applied this method, which has improved the efficiency of reloading startup physical tests and reduced the amount of radioactive wastewater generated in the primary circuit during the test process

The BIPR-8A executable module is a main component of the BIPR 2007 code. This module is intended for three-dimensional coarse mesh calculation of the VVER reactor cores. This paper presents the results of validation of the BIPR-8A executable module.
Validation is based on comparison of the calculated results with the operating data of power plants with the VVER-440, VVER-1000 and VVER-1200 reactors, and with the results of calculations obtained by using Monte Carlo codes.
The results of comparison of the following values at HZP power level and operational power level are presented: critical concentration of boric acid, temperature reactivity coefficient, efficiency of the control rods system, relative power distribution of fuel assemblies and 3D power distribution, effective multiplication factor.
The operational data used for comparison were obtained on the units of the following nuclear power plants: Kalinin NPP, Novovoronezh NPP-2, Leningrad NPP-2, Rostov NPP, Kola NPP.
The calculation results obtained by using Monte Carlo codes MCU-PD and MCNP are presented for the calculation models of the following first and stationary fuel cycles: VVER-1000 – ETE benchmark, Tianwan NPP, Rostov NPP, Novovoronezh NPP-2, Hanhikivi NPP and VVER-TOI.

For successful licensing and safe passage of pilot operation of a new type of fuel in the core of an operating VVER-1000 reactor, it is necessary to carry out a preliminary validation of design codes that are routinely used for developing fuel cycles, core safety analysis and calculating maintenance of power unit operation.

Validation matrix of the design codes includes the problem of calculating transition functions of self-powered neutron detector (SPND). The transition function of SPND is a relation between SPND current and average power of six fuel rods that surround SPND in the FA. Incore detectors are the main part of incore neutron flux monitoring system. They allow to estimate power density in the core.

The article presents the main results of validation of the TVS-M spectral code in relation to the calculation of the transition functions of SPND for VVER-1000 with the new type of fuel. Validation of TVS-M was carried out by comparing with the solutions of modeling tests obtained by means of the MCU-PD Monte-Carlo code. A set of modeling tests was developed, including more than 800 cases. Several types of fuel assemblies with different fuel enrichment, burnable absorber content and SPND location were considered. For each type of FA, various states were simulated, differing in fuel burnup (in the range from 0 to 70 MWd/kgHM), coolant temperature, the presence of control rods and other parameters. A comparative analysis of the results obtained by means of the TVS-M and MCU-PD codes is carried out. In addition, the effects of accounting of heterogeneous structure of SPND and rhodium burnup in the emitter on the transition function were evaluated.

The report describes the results of certification of the ATHLET/BIPR-VVER software package (version 1.0) in the Russian Federation. The ATHLET/BIPR-VVER complex was created on the basis of combining, using the interface, the ATHLET system code, version 2.1A_A (certified in 2014) and the BIPR-8 code of three-dimensional reactor core calculation with the BIPR-8KN kinetic module (stationary calculation block is certified in 2010). The ATHLET thermal hydraulics code has been developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS mbH). The BIPR-8 neutron physical code has been developed by National Research Center «Kurchatov Institute». Cooperation GRS and NRC KI to create ATHLET/BIPR-VVER complex continuing since 1993.
ATHLET/BIPR-VVER complex is designed for coupling neutron-physical and thermal-hydraulic calculation of transient and accident processes in reactor systems with VVER reactor of different types. The ATHLET/BIPR-VVER complex is used at NRC KI to analyze design and beyond design basis accidents prior to the stage of the onset of deformation and damage to core elements.
The report describes the stages of complex certification: a validation report preparing, working with appointed experts, additional investigations performing to increase the level of validation of the complex, preparing of the certification passport. Examples of validation calculations are given that simulate experiments at operating power units and experimental heat-hydraulic stands.
The ATHLET/BIPR-VVER complex, version 1.0 was certified on 10.24.2018 by the Expert Committee for software certification at Rostechnadzor (passport No. 455).

The aim of this work was to create a model of the VVER-440/V-230 reactor in the MCNP code, to calculate the induced activity in ex-vessel structural components and to validate obtained results. The internal reactor components, reactor pressure vessel and external shielding of the V1 NPP reactor were examined and the material composition of the individual components was determined on the basis of available documents. At the same time, material vectors of components entering the physical model were determined. A survey of the V1 reactor operating history revealed that after 12 campaigns shielding assemblies were introduced into the core due to the gradual radiation embrittlement of the pressure vessel, thereby reducing the core and reducing the neutron flux density in the sample volume area. On this basis, representative campaigns were set up for given time periods. A model of the V1 reactor at the core level was created, taking into account the specifics of the V-230 type. At the same time, the choice of the method of scatter reduction was taken into account, which was also reflected in the creation of the geometric model. Based on the proposed methodology, the activation of the reactor concrete shaft at the sample volume location was performed using the MCNP5 with the ENDF/B-VII.1 cross section library. The activities of selected radionuclides were determined based on which it was possible to verify the correctness of the results with respect to the measurements of samples performed by Wood Nuclear Slovakia s.r.o. and measurements performed at the UJFI FEI. The conformity of the calculated values with the measured values demonstrates the correctness of the calculation model. In this way, the validation of the MCNP model for calculating the induced activity of the reactor components of the V1 NPP was achieved.

Natural circulation of the coolant is an important component of ensuring safety and core cooling in conditions when forced flowrate is lossed. Unit 1 of Novovoronezh NPP-2 is the leading unit of the new NPP-2006 project with VVER-1200. When commissioning this power unit, tests were conducted for the first time to verify the natural circulation mode, confirming both the development of this mode while simultaneously shutting down all the MCPs, and as the design value of the permissible reactor power level in this mode.
Due to the significance of the obtained experimental data during the commissioning tests and the need for validation of the calculation model for NPP-2006 project which is used in the calculations for the safety justification, a numerical and experimental study of the natural circulation mode was carried out. This study was jointly conducted by experts from the filial “Novovoronezhatomtechenergo” of the JSC “Atomtechenergo” and the National Research Center “Kurchatov Institute”, taking into account their specializations. The basis of this study was a comparison and analysis of experimental and calculated data.
The main study results are presented. Conclusions and suggestions on the practical use of these results are given.

tWO Generation III+ VVER-1200 units are planning to be built in Hungary. Investigation of core improvements made for the new reactor type has been done by the vendors however appropriate independent calculation tools are advantageous for the users, among others for strengthening the safety level.
Recently several new VVER reactors have been commissioned; among them three VVER-1200 units. To determine experimentally neutron-physics characteristics of those units almost similar start up test programs were performed by extending them with some new, unconventional measurements, too. Several experiments from the physical start-up of a VVER-1200 were published recently.
This contribution summarizes experience gathered at our institute (MTA EK) with our newly developed KARATE-1200 code system on the basis of the reactivity measurements gained from the literature.

1 NRC “Kurchatov Institute”
2 CNNC Jiangsu Nuclear Power Cooperation

During the commissioning of Unit 4 of Tianwan NPP at the power level of about 50% the process of insertion of regulating group 10 in course of boron dilution was performed to measure the group efficiency. Experimental data of the process, including temperatures, flow rates, pressures, control rods positions, in-core and ex-core detectors readings, etc. were carefully recorded by monitoring, control and diagnostic systems. It was then found these data are suitable for modeling by means of coupled three dimensional neutronics and thrmo-hydrolics code.

The paper presents results of preliminary modeling of that process by means of BIPR-ATHLET-VVER code. The problem of transforming of this problem to the benchmark is discussed.

The introduction of new types of nuclear fuel on Ukrainian NPPs for the improvement of operational characteristics and diversification of nuclear fuel supplier is an on-going process. Significant efforts have 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 the introduction of new fuel assembly (FA) types. On first step of the independent analysis of core characteristics, the preparation of cross-section (XS) libraries, the HELIOS code is applied. To ensure this capability, HELIOS models for the actual types of fuel were developed and tested within the framework of BMU/HZDR/SSTC NRS cooperation. The following fuel types are operated now or considered for near perspective for Ukrainian NPPs: TVEL Fuel Company – First and Second Generation and “slim fuel pin (fp)” for VVER-440, TVSA and TVSA-12 for VVER-1000; Westinghouse – TVS-WR (or RWFA) and FA with IFBA fuel for VVER-1000, “ESSANUF” FA for VVER 440.
The result of HELIOS models been developed for new types of fuel and their testing are presented in this paper.

At Loviisa NPP, nodal code HEXBU-3D/MOD5 is utilized in designing the loading patterns as well as in the online monitoring of the reactor core performance. The HEXBU-3D code is based on two-group diffusion formalism and uses parametrized group constants as an input data. In the MOD5 version of the code, the feedback effects are modelled using second-order polynomials.

The current practice to create a group constant library for HEXBU-3D is to apply a miscellaneous set of various programs and scripts that carry out lattice calculation using CASMO-4E and, thereafter, post-process the obtained results into the required format. Improvement in this procedure was seen necessary and, therefore, a small-scale internal R&D program was started in 2019. The target of this project is to develop a completely new tool, XS-TOOL, for generation of group constant libraries for HEXBU-3D.

In this paper, the current status of the XS-TOOL development is documented. The code, written in Python language, is structured into modules including a general-purpose module for handling CASMO’s .cax output file. During the development work, special attention is paid on QA issues, e.g. source code style conventions and version handling, and traceability of the results created using the tool. The XS-TOOL software also includes features that enable the tool to be used for various data analysis purposes according to individual needs.

The paper documents the comparison results confirming that the XS-TOOL produces essentially identical results with the current production tools. The paper also demonstrates results of a spin-off study where limitations of the second-order feedback model were examined. Future plan is to expand the capability of the XS-TOOL for generation of MOD6 format library. In the MOD6 version of the HEXBU-3D code, the feedback effects are modelled using high-order polynomials making the MOD6 version suitable for dynamic analyses.

As long as VVER-440 reactors exist, their fuel design is subject to ongoing improvement. Efforts to find the optimal fuel design continue and encompass increasingly extending ranges of both geometry and material parameters of fuel assemblies.
This paper presents the results of optimized radial enrichment profiling for RK3+ fuel assemblies and discusses the fundamental importance of using burnable absorbers (Gd2О3 rods) in fuel assemblies suggested for prospective fuel cycles.
Regarding RK3+ and RK2 fuel assemblies, this paper also presents thermohydraulic data yielded by computations performed using SC-1 certified software for a VVER-440 seven-assembly core fragment. These computations show that – compared to RK2 assemblies of the same capacity – RK3+ assemblies with optimized enrichment profile have more uniform radial temperature distributions and lower maximum outlet temperatures.

Different calculational tools are used in GRS to perform neutronics analyses. In particular, the neutronics a VVER-1000 core based on the Kalinin-3 benchmark specifications was analysed with DYN3D and PARCS, with the sets of few-group macroscopic cross-sections (XS) generated with SCALE and Serpent, respectively. Although the problem statements and extent of the analyses were different, it is desirable to have possibility to compare the common intermediate step – the generated XS. This is important to provide additional insight to the intermediate results and to identify inconsistencies in the underlying models.

The direct quantitative comparison of XS, as generated with SCALE and Serpent is not possible, since the prepared XS have different energy group structure and were prepared for different sets of parameters. However, both codes provide information about the flux spectra in the output files. The impact to the neutron spectrum of a parameter should have at least qualitatively the same behavior. This work provides the comparison of the changes to the neutron flux introduced by FA type (U enrichment and presence of burnable absorber), fuel temperature, moderator density and boron concentration, as predicted with Serpent and SCALE.

The implementation of the Low Leakage 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. There is presented and discussed a new correction function formulation used during the criticality achievement process assessment.

VVER-440 reactors have been utilized in Slovakia since 1978. Up to now, the length of their operating cycle has not change. This paper presents a proposal of extending the cycle from 12 to 14 months. Main attention is paid to the selection of the equilibrium fuel cycle. In this selection, the current fuel was considered with respect to relevant limits.

Review of AER Working Group B Meeting in Pilsen, Czech Republic is given.

4 units of Paks NPP operate with 15m cycle since 2016. 4.7 and 4.2% enriched fresh fuel assemblies are used and the cycle length is regulated by application of differently enriched fresh fuel. There is a project in progress to verify safety of optimised fuel for Paks, fuel with reduced pin outer diameter and sold pellets. The paper gives some outlook of project status, the plans of transient cycles of Paks NPP and main behaviour of core containing optimised fuel.

Work presents test safety assessment calculation technique for specified operating high power VVER unit in daily maneuvering regimes. Maneuvering operation test is actual problem of VVER reactors exploitation. Such tests are classified as nuclear hazardous procedures which are accompanied by power variations, regulators actions and continuous space power field fluctuations. As a result a lot of possible initial reactor plant conditions are occurred. This fact is taking into account during experiment safety assessment.
Proposed technique bases on conservative approach. Report contains step-by-step method structure, calculation procedures and applied models description for each step. Initial conditions for some RIA accidents are given and justified. Practically significant conclusions are supported by method application results analysis for assigned task. All calculations are implemented by KORSAR/GP programming code with 3D-neutrion kinetic model

Time dependent leakage model
to identify defective fuel assemblies in VVER-type nuclear reactors

Dr. Imre Szalóki1, Gábor Radócz1, Dr. Tamás Pintér2,
Péter Jakab Rozmanitz3, Ilona Varjúné Baracska3, Dr. Anita Gerényi1

1 Institute of Nuclear Techniques, Budapest University of Technology and Economics, Budapest, Hungary
2 Paks II. Nuclear Power Plant Ltd., Hungary
3 MVM Paks Nuclear Power Plant Ltd. (PA Zrt.), Hungary

In order to minimize the radioactive contamination in the coolant loops and in the reactor core of nuclear reactors it needs to be available an effective procedure and spectroscopic hardware systems for perception of leakage fuel. Leaking state of the fuel assemblies are indicated by the level of specific radioactivity’s of the fission products, the transuranium radioisotopes and the time-dependent behaviour of these radioactivity’s in the primary coolant. The characteristic parameters of the time-dependent leakage process can be calculated from the specific activities and the ratios of the radioactivity of iodine isotopes such as 131I, 132I, 134I, 134I, 135I, while the ratio of radioactivity of the isotopes 134Cs, 137Cs can be used to predict the burnup level. Several leakage models have been developed to evaluate the integrity state of fuel assemblies over the last two to three decades. Most of these models are based on calculations from activity measurement at equilibrium state of the reactor. Due to the stationary model is only applicable to equilibrium state, it cannot provide a meaningful description of leakage processes in non-stationary (dynamic) situations, for example spiking and time dependent leakage phenomenon. In the recent years, in an R+D project, we have developed a new calculation model based on the operating conditions of Paks Nuclear Power Plant reactors and on radioactivity data of leaked fission products. These activity data were originated from an automated gamma spectrometric measuring system operating an 8-hour time resolution.
The core of the leakage-model consists of a system of differential equations that describes most of the processes of the formation of fission products and activation events both in the fuel and in the primary coolant. Moreover, the dynamic model considers the surface contamination of the fuel assemblies, radioactive decay processes in the coolant and in the fuel, describes the spiking events as well. The activities of isotopes in the coolant depend on the technological parameters of the reactors such as intensity and efficiency of the purification devices, power of the reactor, those are considered in the model. Recently, by solution of this dynamic model combining with the stationary leakage calculations, the criterion factor and variation of the number of leaking fuel assemblies are determined over time in the reactor campaigns. Reaction rates of neutron-induced activation processes in fuel elements and in the primary coolant were determined by burnup calculation using the MCNP6 program.
Acknowledgments
This work has been partly carried out in the frame of projects supported by MVM Nuclear Power Plant Paks, Hungarian Atomic Energy Authority and VKSZ-14-1-2015-0021 Hungarian project supported by the National Research, Development and Innovation Fund.

Beta emission type of neutron detectors also called self-powered neutron detectors (SPND) are the basis of the in-core instrumentation system ICIS-M. SPND readings is quickly processed and displayed to the operator as an energy release in the all volume of the reactor core. At present, ICIS-M is used at 20 power units with VVER-1000 reactors, and at 3 power units with VVER-1200 reactors.
Algorithms for the transition from SPND currents to the linear power of the all volume of reactor core were developed on the basis of extensive operational experience and numerous measurements. At present, the coefficients for the transition function from the SPND currents to the linear energy release are prepared using the multi-group spectral program TVS-M for each type of fuel assembly and each set of in-core instrumental detector assembly. At the National Research Center “Kurchatov Institute” conducted series of experiments with the irradiation of the SPND at the critical stand which energy spectrum is very close to the VVER spectrum. It confirmed the correctness of the algorithms and programs for calculating functions of the transition from the SPND currents to linear energy release. An indirect confirmation of the accuracy and reliability of the transition from the SPND current to linear energy release during operation of the VVER reactor core is the coincidence of the core power determined from the SPND readings with the power determined by thermal methods.
The transition from linear energy release in measured prisms of fuel assemblies to linear energy release in all calculated prisms of the reactor core is performed using the “Hortitsa‑M” software package, which consists of a power distribution reconstruction module and pin‑wise power distribution calculation module.
Power distribution reconstruction module based on the neutron‑physical model of project code BIPR‑7A and use cross‑section library of this code. The model implemented solution of the 3D two‑group neutron diffusion equation. Pin‑wise power distribution calculation module based on the neutron‑physical model of project code PERMAK-A and use cross‑section library of this code. The model implemented solution of the four‑group neutron diffusion equation.

Starting with 1972, twelve new VVER-440 reactor units were started to build on the territory of Czechoslovakia. Nowadays the 2 oldest units were already shutdown in Slovakia, 4+4 units are still operating at the territory of Czech and Slovak republic, and 2 units of VVER-440 V213 type are still under construction in Mochovce, in Slovakia.

To keep the unit’s operation safe and reliable, wide range of modifications and upgrades were performed. In scope of strengthening the reactor’s core monitoring and surveillance, at Dukovany NPP (1998) and at Bohunice NPP (2001) the original Russian VK3 computation system was completely replaced with an alternative advanced Core Monitoring and Surveillance System SCORPIO- VVER.

Nowadays the SCORPIO-VVER CMS presents a nuclear fuel type and fuel vendor independent, advanced reactor core monitoring system with an open and flexible framework, including the latest achievements in the fields of N/F and T/H for reliable and safe reactor operation with high efficiency of fuel cycle, and governing the know-how and knowledge of 5 European institutes with proven experiences with reactor operation.

Since the first installation the SCORPIO-VVER CMS system has a remarkable operating history and experience. More than 20 years of experiences at 6 units of VVER-440 type of reactors in two different countries helps to put the system to a very high level of usability and reliability. Even the SCORPIO-VVER is installed only on VVER-440 reactors, the development of VVER-1000 version is on the way.

During the upcoming period the system will be prepared to fully support the latest and most advanced PK3+ fuel assemblies loaded into Dukovany NPP reactors in 2023.

Review of AER Working Group C Meeting is given.

Currently, due to the rapid pace of development of computer technology, the development of a multiphysics approach to modeling physical processes is becoming more widespread. Taking into account several physical phenomena and their influence on each other allows you to get more close to real results. Since feedbacks sometimes have a significant impact on the calculation results, the creation of coupled codes is an important component of the development of an approach to modeling physical processes in both nuclear reactors and other fields. In the field of modeling of the nuclear reactors, the consideration of feedbacks between thermal hydraulics and neutron physics is of the greatest importance. To simulate dynamic processes, as a rule, coupled solutions of fast diffusion neutron-physical codes and thermohydraulic sub-channel or best-estimate codes are used. In this paper, we consider the creation of a coupling scheme of the SKETCH-N nodal neutron-physical code and the best-estimate thermohydraulic code ATHLET. Various possible options for exchanging data between codes are considered, after which it was concluded that it is advisable to exchange data between codes by using computer memory (using the MPI library). Verification and validation were performed using the results of the Kalinin-3 international benchmark test. The calculation results and their comparison with experimental data and calculations using other codes show good convergence. During the creation of the coupling scheme, the creation of a system for processing and visualization of calculated data was also begun. Calculations of the stationary state were carried out using both simple (163 channels) and complex (simulation of the complete primary circuit) thermohydraulic models. The calculations of the dynamic shutdown of the MCP were carried out using a complex thermohydraulic model taking into account mass transfer between the channels.

New method of maneuvering for WWER and PWR has been patented (RU 2 675 380 C1 from May 15, 2018 year). It is based on regulation of coolant flow-rate through the reactor, as a new effective and safe means of impact on reactivity. In particular, changing the flow-rate while maintaining a constant value of heating of the coolant in the reactor provides the big part of the required power change. This leads to many favorable results, which have the novelty signs, and advantages over the known maneuvering methods: prevents the occurrence of axial xenon oscillations during maneuvering, provides the most often demanded maneuvering graphs (including primary and secondary frequency control) without the control rods movement and without changing of the boron concentration in the coolant, provides reduced jumps in temperature and pressure in the primary and secondary circles, and reduced jumps of power in the fuel, as well as a number of additional technical and economic advantages. The new technology is justified by calculations using the BIPR-7 reactor code. The correctness of the obtained principal results and conclusions is confirmed by test calculations using the independent system coupled code KORSAR (3D neutron kinetics + Thermo-hydraulics).

Passive ECCS (emergency cooling core system) is one of the most important system of VVER reactors. KORSAR / GP software can use different calculation models for the same elements. In ECCS system different models can be used for modeling of ECCS volume and pipelines. ACCUM element (“hydraulic accumulator”) was developed exactly as ECCS volume. The model of the ACCUM element was verified based on the results of experiments on integral thermo-hydraulic stands. This model has been used during very long time in real safety analysis. The latest versions of the KORSAR / GP software package make it possible to use the SLVES (pressure vessel) calculation element. The mathematical model of SLVES includes a wider list of physical phenomena, to calculate the volume of the ECCS. This work is devoted to testing various models of the passive part of the ECCS.

Today’s computational technology and its rapid development more and more enable the application of higher-order transport approximations and advanced numerical solution techniques even in reactor dynamic calculations. In view of this, a coupled thermal-hydraulics and reactor physics code system is being developed at the Institute of Nuclear Techniques of the Budapest University of Technology and Economics based on a higher-order transport approximation, the simplified spherical harmonics theory. The advantage of this method is that with a small increase in computational effort, it provides additional accuracy compared to diffusion theory. Besides – due to the fact that the multigroup SP3 and diffusion equations have a mathematically similar form – it requires minimal effort to implement an SP3 solution algorithm to an existing diffusion code.

This paper focuses on the developed algorithm, which applies Galerkin weighted residual method for spatial and implicit integration (the backward-difference method) for time discretization. Besides the developed solution algorithm, results of two-group kinetic SP3 calculations are also presented for various one-dimensional perturbations taking into account the delayed neutron precursor balance equations as well.

The nature of the SP3 equations detailed above and the flexibility of the applied finite element algorithm make the developed code named SPNDYN a good starting point for more realistic dynamic calculations in the future.

In the paper the results of comparison of experimental and calculation data for two transients in the VVER-1000 and VVER-1200 cores are considered. The calculations were performed using the code NOSTRA. The code NOSTRA is intended for simulation of processes in reactors VVER cores with defined coolant parameters at the inlet of the core.
The first of considered processes took place during switching-off one of four MCP operating on Unit 3 of Kalinin NPP in the beginning of the first fuel cycle. This process is considered in OECD benchmark. In the presented paper the data from solutions some of tasks of this benchmark are shown.
In addition to deterministic calculation of considered process the statistical analysis of the process with taking into account uncertainties of values of some process parameters and possible errors of the model of the code NOSTRA was performed. The comparison of calculation results using the code NOSTRA not only with experimental data but also with calculation results using other codes is shown.
The second considered process took place in the reactor VVER-1200 core on Unit 1 of Novovoronezh NPP-2 during the switching on of one MCP when the other three MCPs are in operation. This experiment was held at the power ascension stage to 100 %. The initial data on temperature and flow rate distribution in FA was prepared using the certified code ATHLET/BIPR-VVER.
The purpose of given research is the verification of the code NOSTRA on dynamic experiments that are concerned with changing of coolant flow rate. It is shown that the code NOSTRA describes considered processes satisfactory.

For start-up of new types of reactors wide series of experiments are performed. The purpose of such experiments is to obtain experimental data that allow to estimate real values of some neutronic parameters of the core. In particularly, during the start-up of VVER-1200 reactor the experiments of initiating axial and diametral xenon oscillations were performed with help of moving CR CPS group and specific CR CPS correspondingly. After CR CPS had been moved the damped oscillations of spatial power distribution were observed. Spatial power distribution came gradually into equilibrium state due to negative feedbacks. The course of these oscillations, their amplitudes, damping decrements, periods, phases depend primarily on the values of reactivity coefficients from fuel temperature and of moderator density, as well as on the cross section of xenon-135, the effectiveness of the SC RC, the geometry of the core and a number of other factors. The parameters of the core model, which ensure the conformity of the calculated and experimental data on the oscillations, are taken as their experimental values. The values of parameters of the core model that ensure conformity of calculated and experimental data on oscillations are considered as close to experimental data.
Since there are a lot of significant model parameters,the attempt of estimation all the significant parameters of the core for this process on the basis of one experiment often leads to degeneracy of task. In this case the conformity of calculated and experimental data may take place for different sets of calculated model parameters. That’s why the attempt of estimation of model parameters on the basis of two experiments (initiating axial and diametral xenon oscillations) was performed. Such approach is also effective due to reactivity coefficients (in particularly, a reactivity coefficient by coolant density) have different effects on the course of these processes.
For the estimation of probability density function Bayes method was used. The Bayes method was implemented in DRAM method. Priori information for unknown parameters was formed on the basis of their estimations. These estimations were obtained during similar experiments on other VVER reactors. The estimation of error in experimental data was based on the estimation of measurements error in VVER-1200 reactor.
During the construction of Markov chain the procedure of multidimensional interpolation on non-uniform mesh was used. For this purpose the direct simulation of series of processes under consideration by the code of mutual neutronic and thermophysical calculations NOSTRA had been performed before calculations by DRAM method was performed. Further, the series of direct calculations through equal number of steps of Markov chain were added with new calculations. This ensured the high accuracy of interpolation procedure in the point of maximum of probability density function.
In the paper the most probable values of six unknown model parameters of the core including reactivity coefficients, and twodimensional cuts for probability density function for these parameters provided.

One of the determining accidents in the safety analysis of VVER NPP is a main steam line break accident. For this accident the several typical stages of transient can be picked-out. At the first, initial stage, as a result of the initial event the secondary steam-water mixture outflowing into the «leak» begins, that results in the quick secondary pressure decrease and increase in the heat flux removed from the primary circuit. As a result, the coolant temperature of the core inlet decreases and causes an power increase due to the action of feedbacks. At this stage of the accident the maximum values are reached for the fuel temperature and specific enthalpy. The scram brings the reactor into subcritical state.
In this work the first stage of the accident was simulated. The maximum values of fuel temperature and enthalpy were estimated using the pin-by-pin model of fuel assembly. The results were compared with the results obtained using of the «hot channel» model. The calculations were performed for different initial state of the power plant. The realistic and conservative initial state of RP is considered. Recommendations of the using fuel assembly pin-by-pin model for this accident were developed.

Helmholtz-Zentrum Dresden-Rossendorf
Institute of Resource Ecology
Bautzner Landstrasse 400, D-01328 Dresden, Germany
S.Kliem@hzdr.de

The AER Working Group D on VVER reactor safety analysis held its 28th meeting in Garching, Germany, during the period 26-27 June, 2019. The meeting was hosted by GRS Garching and was held in conjunction with the second workshop on multi-physics MPMV-2 and the first workshop on the ROSTOV-2 benchmark. Altogether 20 participants from eleven AER member organizations and seven guests 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 12 presentations and the discussions can be attributed to the following topics:

• Safety analyses methods and results
• Code development and benchmarking
• Future 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.

A new OECD/NEA benchmark entitled “Reactivity compensation with diluted boron by stepwise insertion of control rod cluster” is starting. This benchmark, based on high quality measurements performed at the NPP Rostov Unit 2, aims to validate and assess high fidelity multi-physics simulation code capabilities. The Benchmark is divided in two phases: assembly wise and pin-by-pin resolution of steady-state and transient multi-physics problems.
Multi-physics simulation requires the generation of few-group cross-sections. This task used to be done with deterministic lattice codes, but in the past few years the Monte-Carlo code SERPENT has demonstrate its ability to generate accurate few-group homogenized cross-section without approximations on the geometry.
Since whole core SERPENT models for production of such cross-section libraries would be computationally costly, models of each sub-assembly type in infinite radial lattice configurations have been created.
These cross-sections are then used to evaluate effective multiplication factors with PARCS for different core configurations. The results are compared with the reference SERPENT calculations. In the next step, an ATHLET (GRS in-house thermal-hydraulic system code) model with an assembly wise description of the core (i.e. one channel per fuel assembly) has been developed for coupled PARCS/ATHLET transients test calculations.
This paper describes in detail the models and techniques used for the generation of the few-group parameterized cross section libraries, the PARCS model and the ATHLET model. Additionally, transient simulations are presented, and their results are thoroughly analysed.

The RELAP5-3D© computational system code includes several significant quality improvements over the “standard” versions of RELAP5 program. The most important of these are three-dimensional flow modeling capabilities and the incorporation of three-dimensional neutronic model for description of reactor kinetics. This paper deals with a detailed description of the thermohydraulic and neutronic model of the VVER-1000/320 reactor. Attention is paid to the preparation of neutronic XS data and their arrangement into the input deck of the system code. Earlier versions of RELAP5-3D© included the USER option, which allowed the preparation of neutronic XS data in the form of separate input files that communicated with RELAP5-3D© using external procedures created by the user. This is no longer possible in the latest RELAP5-3D© versions. Now, all the neutronic XS data are incorporated into the general input data file. Their structure follows a well-defined format, in our case it is a GEN-type format. The stationary tests of the reactor model were performed using the latest and older versions of the code. The results are compared, and the differences are discussed. Methodological problems that still need to be overcome are also mentioned.

This paper presents a validation of multiphase CFD code Ansys CFX 18 on selected experiments with subcooled boiling flow. Ansys CFX code can simulate multiphase flow by solving three balance equations for each phase (Euler-Euler approach). Boiling on the heated wall can be simulated with the model proposed by Kurul and Podowski. A set of mathematical models of physical phenomena in boiling bubbly flow was selected and tested on the following experiments with subcooled boiling: DEBORA experiments, ASU experiments, Bartolomej et al. experiment and FRIGG experiment. The DEBORA experiment is a vertical pipe with heated wall. The working fluid is freon R-12. The ASU boiling experiment is a vertical annular channel with heated inner wall and freon R-113. The Bartolomej experiment is a vertical heated pipe with boiling water. The FRIGG experiment is a vertical channel with 5 heated rods and boiling water. It was found out that CFX code can reasonably reproduce measured data in all these experiments. At the end of this paper we present an application of this modelling approach to the real case with a complicated geometry: simulation of subcooled boiling in the VVER-1000 fuel assembly with bent rods. The 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“.

Turbulent heat transfer through gas cooled systems is a key factor for the improvements of the Gas cooled Fast Reactors (GFR). Within the FP7 European project THINS (Thermal Hydraulics of Innovative Nuclear Systems), experimental and numerical tools for investigations of the thermal hydraulics of next generation rector systems were developed, applied and validated for innovative coolants. One of the tools is the L-STAR facility, which has been designed and erected at the Karlsruhe Institute of Technology (KIT) to study turbulent flow behaviour and heat transfer enhancement characteristics in gas cooled annular channels under a wide range of conditions. The test section consists of an annular hexagonal cross section channel with an inner electrical heater rod element, placed concentrically within the test section. This design represents the flow domain around a single fuel rod in a future GFR.
In this paper simulation results for the fluid flow at uniform heat release conditions for the smooth heater rod are presented. Application of the facility allows using various values of heating power of the heated rod and fluid mass flow depending on the chosen scenario. The purpose of this study is to create CFD (Computational Fluid Dynamics) models for the L-STAR that are capable of reproducing the results of reference measurements.
The calculations were carried out with ANSYS CFX 17 code. All simulations were performed as steady-state calculations. Mesh, structural element and turbulence model sensitivity studies were also performed during this work. With these model results it becomes possible to validate other CFD models to be applied for the ALLEGRO experimental gas cooled reactor fuel assemblies.

Review of recent AER Working Group G (Thermal-hydraulics and CFD) meetings will be presented.

1: UJV Rez a. s., Dept. of Safety Analyses, Hlavni 130, 250 68 Husinec-Rez, Czech Republic
2: Czech technical University in Prague, Faculty of Mechanical Engineering, Technicka 4,
166 07 Praha 6, Czech Republic

The goal of this work was to evaluate crossflows in a mixed core of VVER reactor. The mixed core consists of old and new fuel assemblies with different geometry. These fuel assemblies have different spacer grids and mixing grids at different elevations. The old and new fuel assemblies also have slightly different total pressure loss. This causes crossflows between neighboring fuel assemblies. Crossflows were evaluated with VIPRE-01 subchannel code and with Ansys Fluent CFD code. In both codes, the computational domain covers an interface among fuel assemblies of different type along the whole core height. VIPRE-01 domain is modelled pin-by-pin and it includes corner plates. Spacer and mixing grids are represented by local pressure losses. CFD model has much finer grid resolution. Spacer and mixing grids are modelled by thin walls and porous zones. Due to the enormous size and complexity of the CFD domain, the mixing vanes were not modelled. Transverse coolant velocities along the core height were calculated in both codes and compared. The results obtained with subchannel code are in a good agreement with results from CFD simulation.

The paper studies the potential for a self-sustaining chain nuclear fission reaction during the progress of a severe accident at Ukrainian nuclear power plants with VVER-440 reactors. The Rivne NPP unit 2 core was modelled. Some models for criticality calculation at different stages of a severe accident in the VVER-440 pressure vessel were developed and corium multiplication properties were calculated.
The severe accident in the VVER-440 core was divided into eight 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. In view of criticality the supply of water (re-flooding) as a part of accident management is accounted. A compensatory measure such as addition of the boric acid solution to the cooling water was analyzed.
The results obtained in the study show whether the fuel rods remain intact at the beginning of a severe accident or whether criticality might appear. With further progression 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, criticality may occur.

This article is devoted to an optimization scenario of the spent nuclear fuel storage pool located at NPP Mochovce site, which, due to safety and legislation, has 71 empty positions (without fuel assemblies). The calculation was performed using MCNP5 and Serpent 2 codes using the nuclear section library ENDF / B-VII. The safety analysis was performed on a three-dimensional model of the storage pool loaded with VVER-440 W-type fuel assemblies with a nominal average enrichment of 4.87 wt.%. The sensitivity of the multiplication factor to the change in the lattice spacing of the fuel rods, the pitch of the fuel assembly, the temperature of the coolant, the density of the fuel and the mass of boron in ATABOR absorbing steel is analyzed. Parametric analysis showed that the propagating properties of the system are negatively affected by an increase in the pitch of the lattice of the fuel rods, an increase in the fuel density, a decrease in the pitch of the absorbing pipes, a decrease in the temperature of the coolant and a decrease in the boron content in the absorbing pipes. Then, the results of the parametric analysiss are used in the spent fuel storage model to propose an optimized scenario for loading the storage using absorbing parts of the VVER-440 control assemblies – so the storage can be fully loaded. In terms of the efficiency of using the maximum number of empty positions in the storage pool and the minimum number of absorber parts loaded, the scenario K_M3-09, in which only 9 pieces of absorbers are used.

This work is focused on criticality safety analysis of spent fuel storage pool for the NPP Mochovce 12. The MCNP5 code with the ENDF/B-VII libraries were applied to model the compact grid of the storage pool loaded with 4.87% enriched fuel assemblies. In order to increase the operational storage capacity of the compact grid, number of absorber assemblies were inserted into the pool instead of four empty rows now needed to reach the required subcriticality. Conservative approach was choosen and applied on parameters and simplifications. The methodology is focused on determination of the best geometrical configuration of the inserted absorber parts in the compact grid. The pre-calculations with simplified model were used to lower the calculation time which was affected by complexity of the model and calculation parameters. The best resulting scenarios were later used in more complex conservative MCNP model with higher precision and for optimalization of criticality safety analysis.

UNCERTAINTIES ASSOCIATED WITH NUCLEAR DATA WHEN CALCULATING FUEL ISOTOPE CONCENTRATIONS AT BURNUP

Pavel Mikoláš
ŠKODA JS a.s., Orlík 266/15, 316 00 Pilsen, Czech Republic

The programs used to determine isotope concentrations in spent nuclear fuel use a multi-group transport approach. Cross-sections and other variables are prepared for multi-group libraries for these programs, which are based on basic nuclear data. These nuclear data cannot be determined accurately, they always have some uncertainty.
The paper shows, on a concrete example, the effect of these uncertainties on the determination of concentrations of selected isotopes in spent fuel by comparison with experimentally determined values of these concentrations.
It is shown that the effect of nuclear data uncertainties is different for different isotopes, and in some cases it is quite considerable.

Topic: Spent fuel disposal and actinide transmutation
The interest in impurities and activation of structural elements is driven by requirements related and imposed by studies connected to deep geological repositories (DGR). The advances in HELIOS2 and the significant extensions in is depletion chains and supported isotopes allowed the develop-ment and implementation of such model in the Studsvik’s SNF code. Preliminary numerical results based on measured and theoretical impurity contents are presented in this paper. The results from the SNF codes are comparted to an explicit HELIOS2 model.

Review of AER Working Group E Meeting in Modra Harmónia, Slovakia is given.

VUJE Inc., Okružná 5, SK 918 64 Trnava, Slovakia

Experimental neutronic benchmark at fast spectrum was defined based on experimental data from sodium cooled fast reactor CEFR exploited at China. It was accepted by IAEA at Vienna as CRP-I31032 On Neutronics Benchmark of CEFR Start-Up Tests. Solution of the benchmark is under way in VUJE. First part – blind solution is characterised at this paper including partial results.

Regular 20th session of the AER Working Group F «Nuclear Fuel Cycle Perspectives and Sustainability» was organized by National Research Centre “Kurchatov Institute” held in Kurchatov sq.,1, Moscow, Russian Federation, July 22 – 25, 2019. 25 specialists from 4 organizations from 2 countries participated in the preparation of 13 reports submitted to the group F. These reports cover a variety of topics, including:
1. Advancement of Nuclear Power structure and technologies for sustainable development.
2. The choice of strategy for development of two-component (ТR and FR) Nuclear Energy System with closed nuclear fuel cycle.
3. Keeping two-component Nuclear Energy System supplied with fuel during TR and FR lifetime taking into account regenerated plutonium sharing.
4. REMIX- and MOX-technologies for VVER operating in closed nuclear fuel cycle.
5. Methods for regenerated uranium isotopic composition correction.
6. Heavy nuclides employment as burnable absorbers.
7. Thorium involvement in VVER reactors fuel cycle.
8. VVER with fuel self-supporting.
9. Incineration of MA and transmutation of long-lived RAW in critical and subcritical reactors.
10. Requirements to RAW handling technologies.
Working group F has consolidated the understanding that
– minimization of stored spent nuclear fuel (SNF) volume,
– gradual transfer nuclear energy system (NES) to the fuel self sufficiency mode (reliable supply with fuel through fissile isotopes breeding from U-238 and Th-232)
can be achieved by
– closing the nuclear fuel cycle (NFC),
– including fast reactors (FR) in the NES,
– improving fuel use in thermal reactors (TR).
The next 21st meeting of working group F “Prospects and sustainability of the nuclear fuel cycle” is scheduled to be held in Moscow (Russia) in July 2020.

Opennig speech for 29th Symposium of AER

Welcome to 30th Symposium of AER in Russia.

Common photo in the front of ENERGOLAND.

Common photo in the front of ENERGOLAND.

Common photo in the front of ENERGOLAND.

Common photo in the front of ENERGOLAND.

Common photo in the front of ENERGOLAND.