26th Symposium of AER on VVER Reactor Physics and Reactor Safety

INFORMATION ABOUT AER WG A ON IMPROVEMENT, EXTENSION AND VALIDATION OF PARAMETRIZED FEW-GROUP LIBRARIES FOR VVER 440 AND VVER 1000Pavel MikolášE-mail: pavel.mikolas@skoda-js.cz, Tel: +420 378042828 ŠKODA JS a.s., Orlík 266/15, 316 00 Plzeň, Czech RepublicABSTRACTJoint 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 fourth meeting was hosted by VÚJE, a.s. in Modra-Harmónia, Slovakia, during the period of April 18 to 20, 2016.In total 19 participants from 11 member organizations took part in the meeting and 14 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.Presentations were devoted to some aspects of transport and diffusion calculations, and especially to the benchmark dealing with VVER-440 and VVER-1000 core periphery power tilt.A. Scherenko presented paper “The ‘Full-core‘ VVER-1000 benchmark calculation by the codes included into the KASKAD code package and by the MCU code. The proposals for the benchmark specification”, Gy. Hegyi (co-authors A. Keresztúri, Cs. Maráczy and E. Temesvári) spoke about “New solution for the MIDICORE Benchmark by the KARATE code system“, I. Pós (co-authors S. Patai-Szabó and T. Parkó) added presentation “Solution of MIDICORE VVER-1000 benchmark by HELIOS code“, A. Keresztúri spoke about “Special problems of group constant generation for spectrum reactors” and finally M. Žák supplied paper entitled “Uncertainty calculation of fuel pin deflection in VVER-1000 fuel assembly”.During the meeting there were discussed the questions connected with the benchmark solutions and, among others, future activities, which are also shortly described at the end of the paper.A technical tour was also accomplished. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 1 10 – 14 October 2016, Helsinki, Finland

ADVANCES IN HELIOS2 NUCLEAR DATA LIBRARYCharles Wemple, Teo Simeonov Studsvik Scandpower Inc. charles.wemple@studsvik.com teodosi.simeonov@studsvik.comABSTRACTThe ongoing development of the HELIOS2 code system at Studsvik includes periodic updates of the nuclear data library. The library expansion includes an update of the cross section data source to ENDF/B-VIIR1, a significant expansion of the burnup chains, addition of a more complete set of gamma production data, and development of new resonance treatment options. The goal is to provide the capability for HELIOS2 to more accurately model a wider array of reactor applications and enhance interoperability with SNF, the Studsvik spent fuel analysis code. This paper will also provide a discussion of the benchmarking effort and an overview of other HELIOS2 development efforts.KEYWORDS: nuclear data library, cross sections, burnup, ENDF/B-VII, HELIOS, SNF 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 2 10 – 14 October 2016, Helsinki, Finland

SOME UNCERTAINTY RESULTS OBTAINED BY THE STATISTICAL VERSION OF THE KARATE CODE SYSTEMI. Panka, Gy. Hegyi, Cs. Maráczy, E. Temesvári Centre for Energy Research, Hungarian Academy of Sciences Reactor Analysis DepartmentH-1525 Budapest 114, P.O. Box 49, Hungary istvan.panka@energia.mta.huABSTRACTThe survey of the uncertainties has been going on for several years in the frame of the Uncertainty Analysis in Best-Estimate Modeling (UAM) LWR benchmark organized by the OECD NEA. The goal of this benchmark is to determine the uncertainties of the coupled reactor physics/thermal hydraulics LWR calculations at all stages. Having experiences in this field, it was decided to elaborate the so called statistical version of the KARATE code system.In the first part of the paper, the main features of the elaborated new code system are discussed. The applied statistical method is based on Monte-Carlo sampling of the considered input data taking into account mainly the covariance matrices of the XSs and/or the technological uncertainties. Additionally, the sampled input uncertainties are propagated through the full calculation chain.In the second part of the paper, some uncertainty results – obtained by the statistical version of the KARATE code system at nodal level – are discussed for an equilibrium cycle related to a VVER-440 type reactor. This means also that the burnup dependence of the considered output uncertainties – in the term of empirical standard deviations and correlations of some safety related parameters (e.g. effective multiplication factor, boron concentration, assembly-wise radial power and burnup distribution) – are discussed, as well. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 3 10 – 14 October 2016, Helsinki, Finland

ACCOUNTING OF BORON-10 CONTENT CHANGES IN BORIC ACID DURING FUEL BURNUP IN REACTOR VVER-1000Bychkova N.A., Lazarenko A.P., Pavlovichev A.M., Sumarokov M.A., Tomilov M.Y. National Research Centre “Kurchatov Institute”ABSTRACTThe implementation in the complex KASKAD the algorithm of taking into account of changes boron-10 content in boric acid during the fuel burnup calculation in the reactor VVER-1000 is described. Influences of boron-10 depletion and main processes of water exchange on boric acid critical concentration are demonstrated. It is shown that the inclusion of boron-10 depletion improves the agreement between measured and calculated values of boric acid critical concentration in the coolant during operation of the fuel cycle.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 4 10 – 14 October 2016, Helsinki, Finland

ANDREA 2.2 AND 2.3 – ADVANCES IN MODELING OF VVER CORESFrantišek Havlůj, Jonatan Hejzlar, Radim Vočka, Jiří Vysoudil ÚJV Řež a.s., Czech RepublicABSTRACTIn 2016 a new version of code ANDREA for core design and reload safety analysis of VVER reactors has been released. The new code version includes several major improvements. The first of them is a seamless incorporation of short time kinetics calculations (without temperature feedback) into the code. Evaluation of transients starting from any zero power state during the cycle is possible. This new feature accompanied by the possibility of ex-core detector signal predictions enables precise interpretation of dynamic measurements of control assembly weight during the reactor start-up. Second important enhancement resides in new flexible format of cross section libraries and in new fuel temperature model based on results of TRANSURANUS fuel performance code. From the user point of view the code now has only one cross section library covering whole range from cold states to reactor full power operation. Furthermore for the new version 2.3 which is to be released shortly we have implemented the possibility of fluent control assemblies motion and of non-equidistant axial nodalisation schemes in VVER- 440 calculations. New implementation of control assembly movement affects also calculations with control rods, where the cusping effect is eliminated.The new code version has been thoroughly tested and validated for both VVER-440 and VVER- 1000 reactors.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 5 10 – 14 October 2016, Helsinki, Finland

VVER-440 FULL-CORE BENCHMARK BY SPECTRAL CODESPetr DAŘÍLEK, Michal SEČANSKÝ, Amine BOUHADDANE, Radoslav ZAJAC and František ČAJKOVUJE Inc., Okruzna 5, SK 918 64 Trnava, Slovakia Petr.Darilek@vuje.skABSTRACTThe Full-Core benchmark is a two-dimensional calculation benchmark based on the VVER- 440 reactor core cold state geometry with taking into the account the geometry of explicit radial reflector. In this contribution the paper offers the results of SERPENT 2.1.25 code and the preliminary results of HELIOS 2 code. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 6 10 – 14 October 2016, Helsinki, Finland

“FULL-CORE“ VVER-440 EXTENDED CALCULATION BENCHMARKVáclav Krýsl, Pavel Mikoláš, Daniel Sprinzl, Jiří Švarný (Corresponding author: pavel.mikolas@skoda-js.cz, Tel: +420 378042828, Fax: +420 378042305) ŠKODA JS a.s., Orlík 266/15, 316 00 Plzeň, Czech RepublicABSTRACTThe calculation benchmark “Full-Core” VVER-440 was proposed few year ago for macro- codes used for neutron-physics calculations for VVER-440 reactors. This 2D benchmark was used to test the power distribution on the pin by pin level for several macro-codes ¹. Recently, extension of this benchmark has been proposed ². In this case the absorption part of control rod is used instead of fuel part for one selected group. The extension of this benchmark was presented at the 25th Symposium of AER in 2015. The reference solution has been calculated by MCNP code using Monte Carlo method and the results have been published in the AER community. In this paper we will compare the available macro-codes results of this extended calculation benchmark.¹ V . Krýsl et. al.: ‘Full-Core’ VVER-440 calculation benchmark. Kerntechnik, V ol. 79, No. 4, August 2014.² V. Krýsl, P. Mikoláš, D. Sprinzl, J. Švarný.: ‘Full-Core’ VVER-440 benchmark extension calculated by MOBY-DICK macrocode. AER Symposium, Balatongyörök, Hungary, 13.-16. October, 2015. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 7 10 – 14 October 2016, Helsinki, Finland

PROPOSAL OF ‘FULL-CORE’ VVER-1000 CALCULATION BENCHMARKVáclav Krýsl, Pavel Mikoláš, Daniel Sprinzl, Jiří Švarný (Corresponding author: pavel.mikolas@skoda-js.cz, Tel: +420 378042828, Fax: +420 378042305) ŠKODA JS a.s., Orlík 266/15, 316 00 Plzeň, Czech RepublicABSTRACTRecently, 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 geometry. 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. The main task of this benchmark is again to test the pin by pin power distribution in fuel assemblies 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. In this contribution we define the ‘Full-Core’ VVER-1000 calculation benchmark and we present the preliminary reference Monte Carlo calculation results.¹ V. Krýsl et. al.: “Full-Core” VVER-440 calculation benchmark. Kerntechnik, Vol. 79, No. 4, August 2014. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 8 10 – 14 October 2016, Helsinki, Finland

MONTE CARLO CALCULATIONS OF VVER-440 AND VVER-1000 FULL-SCALE CORES BY MCU CODED.S. OleynikNational Research Centre “Kurchatov Institute”, Russia, MoscowABSTRACTBecause of the restricted experimental data of power distribution due to design codes’ verification Monte Carlo codes it is taken advantage of Monte Carlo codes. These codes simulates the interaction of radiation with substance on the basis of the information from files of the evaluated nuclear data (i.e. the most exact data without additional assumptions is used). Besides, this method practically does not impose restrictions on the geometry of considered systems.In the frame of AER group the calculation benchmarks of VVER-440 and VVER-1000 full core are prepared for reliability validation of design codes. These benchmarks are calculated by means of MCU (Monte Carlo Universal) code. The results of the calculations are presented in the paper.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 9 10 – 14 October 2016, Helsinki, Finland

COMPREHENSIVE SOLUTION OF THE „MIDICORE” BENCHMARK BY THE KARATE AND MCNP CODE SYSTEMGy. Hegyi, A. Keresztúri, Cs. Maráczy, I. Panka, E. Temesvári, G. Hordósy Hungarian Academy of Sciences, Centre for Energy ResearchReactor Analysis Department H-1525 Budapest 114, P.O. Box 49, Hungary hegyi.gyorgy@energia.mta.huABSTRACTThe modern 3D steady-state reactor core calculations require the modelling of the reflector, where large differences of neutronics properties between the core and reflector regions can be observed.This problem can be observed in VVER-1000 NPP also, when pin power distribution is determined by codes based on pin to pin diffusion difference method on one side and by codes based on nodal diffusion method with pin power reconstruction on the other side. To study this phenomenon, a mathematical benchmark representing a simplified sector of VVER-1000 core was defined by ŠKODA JS a.s. in cooperation with ÚJV Řež a.s. It consists of 37 fresh fuel assemblies with 4 different enrichments. The details about core basket and the geometry of explicit radial reflector are given, too.To investigate this benchmark is valuable in many reasons. The validation of the actually applied calculational methods is necessary, although the distribution of the fuel pin power is not directly measured.It is worth to see how our models which were used for VVER-440 until now can be applied for VVER-1000.The problem has been solved by MCNP and KARATE codes. Now some improvements will be presented. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 10 10 – 14 October 2016, Helsinki, Finland

AER X2 BENCHMARK DRAFT DECISION WITH USE OF PARCS CODEM.Ieremenko, I.OvdiienkoState Scientific and Technical Centre on Nuclear and Radiation Safety, Kyiv, Ukraine Stusa st. 35-37, 03142 Kyiv, Ukraine yn_ovdienko@sstc.com.uaABSTRACTAt the 19th AER symposium a benchmark on core burnup calculations for VVER-1000 reactors with loadings of TVSA fuel assemblies was proposed for further validation and verification of the reactor physics code systems.This report presents the results of benchmark’s TASK 1 and TASK 2 decision with use of PARCS code for 1st fuel cycle. The given results consist of the 3D core burnup calculation together with calculations of critical states for hot zero power conditions. Also there are solutions for reactivity effect calculations for different reactor states and the analysis of critical boron acid concentrations, 2D power density and burnup distributions.The work was performed in the framework of the project BMUB/GRS 3614R01520 under German BMU support.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 11 10 – 14 October 2016, Helsinki, Finland

THE X2 BENCHMARK FOR VVER-1000 REACTOR CALCULATIONS. OVERVIEW AND CURRENT STATUS.T. Lötsch1), S. Kliem2), E. Bilodid2), V. Khalimonchuk3), A. Kuchin3), Yu. Ovdienko3), M. Ieremenko3), R. Blank4), G. Schultz4)1): TÜV SÜD Industrie Service GmbH, Energy and Systems (IS-ES), Westendst. 199, 80686 Munich, Germany;2) Helmholz Zentrum Dresden-Rossendor, Bautzener Landstr. 100, Dresden, Germany; 3): State Scientific and Technical Centre for Nuclear and Radiation Safety of Ukraine (SSTC N&RS), Stusa st. 35-37, 03142 Kyiv, Ukraine4) IBBS, Ingenieurbüro Blank-Schultz, Berlin-Schildow, GermanyABSTRACTThe paper gives an overview about the tasks defined in the framework of the X2 benchmark, firstly proposed at the 19th symposium of the AER in 2009. The X2 benchmark was proposed for further validation and verification of the reactor physics code systems for VVER-1000 reactors with loadings of TVSA fuel assemblies. The X2 benchmark comprises all stages of steady state and transient reactor calculations starting with the fuel assembly data preparation. Therefore X2 benchmark specifies the FA and core characteristics as well as the core loading patterns of four consecutive burnup cycles for a Ukraine VVER-1000 reactor core. A set of operational data for comparisons with steady state reactor core burnup calculations and transient neutron kinetics calculations were provided. Such a benchmark is useful for validating and verifying the whole system of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. Reviewed and consolidated results of the tasks 1 and 2 for X2 benchmark steady state calculations were presented during the last years at the AER symposia. In the framework of several projects supported by the German BMU5) the 3D neutron kinetic code DYN3D and the coupling of DYN3d with thermohydraulics system codes were further validated and verified on the basis of the data provided in the framework of the X2 benchmark. In preparing results for the X2 benchmark several organisations have been participated: IBBS, HZDR, SSTC, TÜV SÜD. On that basis TÜV SÜD has been provided the analysis and formulation of the specific X2 benchmark tasks. As continuation of the work on the X2 benchmark the tasks were extended with task 3 including data of the 3D calculations results and pin-by-pin distributions for selected fuel assemblies as well as task 4 providing data for 3D neutron kinetic calculations of reactor transients. The paper presents the current state of the X2 benchmark and discusses new results as continuation of the work started with the X2 benchmark proposal in 2009.5) The work was partly performed in the framework of project BMUB 3614R01520-868100/12. The report describes the opinion and view of the contractor – TÜV SÜD Industrie Service GmbH, IS-ET – and does not necessarily represent the opinion of the ordering party – BMUB-BfS/GRS.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 12 10 – 14 October 2016, Helsinki, Finland

CALCULATION OF VVER BENCHMARKS BY CODES INCLUDED INTO SOFTWARE PACKAGE KASKADAleshin S.S, Oleynik D.S., Shcherenko A.I.National Research Centre “Kurchatov Institute”, Russia, MoscowABSTRACTSome years ago the authors from the Check Republic proposed the full core 2-D benchmark of VVER-440. The similar 2-D benchmark related to the full-core of VVER-1000 was proposed at the AER Working Group A&B meeting last year. Such type of benchmarks requires the adequate specification of the volume surrounding the core which geometry is complex enough.This work contains the detailed description of the calculation models of the PERMAK-A code.The proposed benchmarks have been solved using the BIPR-7A and PERMAK-A codes included into the software package KASKAD. The data calculated by the engineering codes have been compared with the calculation results of the Monte-Carlo code MCU. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 13 10 – 14 October 2016, Helsinki, Finland

CALCULATING NEUTRON DOSIMETER ACTIVATION IN VVER-440 SURVEILLANCE CHAINS WITH SERPENTTuomas Viitanen and Jaakko Leppänen VTT Technical Research Centre of Finland tuomas.viitanen@vtt.fiABSTRACTThe structural integrity of reactor pressure vessels (RPVs) can be studied by preparing test specimens from the RPV material, irradiating the specimens in the surveillance position or positions at the reactor periphery and measuring the material properties of the irradiated samples. To match the measurements of the test specimens to the state of the reactor pressure vessel at a certain moment of time, the neutron exposure of the irradiated test specimens as well as the reactor pressure vessel need to be determined.The exposure, more precisely the neutron fluence, is usually calculated using either deterministic or Monte Carlo calculation codes. Accuracy of the computational estimates can be increased by means of neutron dosimetry, i.e. by normalizing or adjusting the computational results to match the measured activation of neutron dosimeters. For this reason, also the surveillance specimens of Loviisa-1 and Loviisa-2 VVER-440 units are always irradiated together with several neutron dosimeters: The axial profile in the neutron fluence is monitored using Fe/Ni dosimeter discs, in addition to which the fluence spectrum is measured using separate wire dosimeters.In the current work, the measured activities from neutron dosimeters irradiated in the surveillance position of Loviisa-1 unit are used to validate Monte Carlo reactor physics code Serpent for calculations at the reactor periphery, for example surveillance position or RPV. The neutron source in the Serpent calculation is generated based on a full-core power distribution from simulator code HEXBU-3D, and Serpent is only used to calculate the neutron transport from the source points in the reactor core to detector locations. Since the neutron flux decreases by orders of magnitude between the reactor core and the locations of interest, the convergence of the Monte Carlo transport solution needs to be accelerated using new weight-window based variance reduction techniques of Serpent 2.1.27. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 14 10 – 14 October 2016, Helsinki, Finland

CALCULATING THE SPECIFIC ACTIVITIES OF VVER-440 DOSIMETERS WITH MAVRIC AND KENOAntti RätyVTT Technical Research Centre of Finland antti.raty@vtt.fiABSTRACTThe structural integrity of reactor pressure vessels (RVP) can be studied by irradiating specimens of RVP material at the reactor periphery and measuring the properties of these samples.The aim of this work is to calculate the neutron fluxes and corresponding reaction rates for certain dosimeter samples at Loviisa power plant. Dosimeter samples are irradiated inside Loviisa RVP at downcomer side relatively close to the core. The calculation model was built with Oak Ridge National Laboratories’ MAVRIC and KENO codes. All the bundles all pressure vessel internals were included in the model. Reactor upper and lower parts were omitted. MAVRIC is based on deterministic methods and enables an efficient variance reduction technique for fast and precise modelling with Monte Carlo code KENO.Public data on VVER-440 core and fuel was used to model the geometry and fuel bundles, but actual fuel enrichment, burnup and power density data were obtained from Fortum. Source spectrum was based on MAVRIC example spectrum for U-235 and Pu-239.Surveillance chains have been irradiated for several power cycles. Total activities for dosimeters were calculated by modelling the core for each of these cycles separately and summing up the activities taking into account decay time after each cycle. In the end, results were compared with measured values. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 15 10 – 14 October 2016, Helsinki, Finland

TESTING OF DYN3D NODAL AND PIN-POWER SIMULATION OF VVER- 1000 MINI-COREJan HádekÚJV Řež, a. s., Hlavní 130, 250 68 Husinec-Řež, Czech Republic jan.hadek@ujv.czABSTRACTThe pin-power reconstruction is a computationally efficient approach how to achieve higher resolution in the neutronic/thermal-hydraulic simulation of reactor core. Generally, the result of heterogeneous (pin-power) reconstruction in individual positions of fuel pins of homogeneous fuel assembly can be expressed as a product of homogeneous reconstruction in three-dimensional reactor dynamic code (sometimes it is called as amplitude function) and pin- by-pin power distribution results obtained with using of lattice code (form function). In this study, DYN3D nodal and pin-power reconstruction simulations were implemented and numerically validated with transport reference solutions for a VVER-1000 mini-core of 7 fuel assemblies. At this stage, the testing was performed at hot zero power state to enable comparison with transport reference solutions received by Tripoli4 code. The homogenized cross-sections library in nemtab format and pin-by-pin power shape library for unrodded and rodded states and for fresh and irradiated states were prepared by Apollo2 code. Computational results are compared and discussed.This work was partially funded by the NURESAFE project (Contract number: 323263) in the 7th Euratom Framework Programme of the European Union. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 16 10 – 14 October 2016, Helsinki, Finland

ADVANCED PHYSICAL STARTUP TEST FOR VVER-1200 OF NOVOVORONEZH NPP: TECHNIQUE AND SOME RESULTSD.A. Afanasiev, Yu.A. Kraynov, A.A. Pinegin, S.V. Tsyganov NRC “Kurchatov Institue”ABSTRACTStartup physics tests intended for confirmation of design characteristics of the core loading and their compliance with safety analysis preconditions. The programme of startup tests for leading unit usually composed in a way to study as much as it possible neutron-physical characteristics in the safest condition of zero power. State-of-the-art safety analysis is now including computer codes that use three dimensional neutron kinetics and thermohydrolics models. For substantiation of such models, for its validation and verification there is a need in reactor experiments that implementing spatially distributed transients.When composing hot zero power physical startup programme for new VVER-1200 unit of Novovoronezh NPP we started from that statements. Several unconventional for VVER tests were developed for that programme. It includes measuring the worth for each of control rod groups and measuring of single rod worth form the inserted groups – test that models rod ejection event in some sense.Physical startup of unit 6 of Novovoronezh NPP with the first VVER-1200 took place in May 2016. During physical startup programme at hot zero power all planned new tests were performed. The presentation discusses new tests technique, its practical implementation, experimental equipment and some results obtained at the first VVER unit of Generation “3+”.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 17 10 – 14 October 2016, Helsinki, Finland

CALCULATIONS OF VVER-1000 START-UP PHYSICS TESTS USING THE MCU MONTE CARLO CODEA.S. Bikeev, M.A. Kalugin, D.A. Shkarovsky, Anna I. Shcherenko, A.A. Pinegin, D.V. AfanasevNRC “Kurchatov Institute”, 123182, Russia, Moscow, Kurchatova sq., 1ABSTRACTCalculations of start-up physics tests for the unit 3 of Rostov NPP were carried out using the MCU Monte Carlo code.For calculations of start-up tests, a full-scale computer model of VVER-1000 was developed using the design documentation for fuel assemblies and a reactor facility: reports, blueprints, explanatory notes, and technical validations. The model consists of a core, a base plate, a baffle, a reactor pit, a reactor vessel, and beyond vessel space, including a dry protection, a concrete shaft, a support girder, a biological protection. The reactor core was assembled from 163 fuel assemblies with fresh fuel, according to the pattern of the load 1 of the unit 3 of Rostov NPP. The special attention was given to the accuracy of the geometry and material description of fuel and control rods, spacer and mixing grids, thimble tubes, bottom and top nozzles during the FA model development. The model of a fuel rod consists of a cladding, fuel pellets, a spring lock, and top and bottom end-pieces.Precision calculations of the following neutron physical characteristics were carried out: the boric acid concentration in coolant for all critical conditions of the core, the efficiency of single control rods, the integral and differential efficiency of the RCCA groups, the efficiency of the emergency protection and the reactivity coefficients.The comparative analysis was made for the calculated and experimental values. As a result, adequate deviations were attained.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 18 10 – 14 October 2016, Helsinki, Finland

AER WORKING GROUP B ACTIVITIES IN 2016Petr DARILEKVUJE Inc., Okruzna 5, SK 918 64 Trnava, Slovakia petr.darilek@vuje.skABSTRACTReview of AER Working Group B Meeting in Modra – Harmónia, Slovakia is given. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 19 10 – 14 October 2016, Helsinki, Finland

RECENT IMPROVEMENTS AND OPERATING EXPERIENCES OF LOVIISA NPP FUEL MANAGEMENTS. Saarinen, J. Kuopanportti, T. Lahtinen, J. Arjoranta Fortum Power & Heat Ltd, Espoo, Finland simo.saarinen@fortum.comABSTRACTLoviisa NPP has two VVER-440 reactors with the thermal power of 1500 MW. There have been several changes impacting the fuel economy of Loviisa reloads during the past 10 years. Ten years ago, the plant was operated with 3 year loading scheme with fuel assembly average enrichment between 3.7 % and 4.0 %. In 2009 first Gd fuel assemblies with higher enrichment, namely 4.37 %, were loaded into Loviisa. The enrichment increase enabled to move to a 4 year loading scheme. In connection with the transition to the 4 year loading scheme, the assembly burnup limit was increased from 45 MWd/kgU to 57 MWd/kgU.The fuel economy in the 4 year loading scheme has been further improved by introducing mixing vanes to the spacer grids. The mixing vanes improve the internal coolant mixing inside the fuel assembly. Thus the with same pin powers, the maximum subchannel outlet temperature remains lower with the mixing vanes than without them.The engineering safety factors for Loviisa were reassessed. The reassessment lead to a significant reduction of the subchannel enthalpy engineering safety factor from 1.16 to 1.10. With this change, a fully low leaking loading pattern was possible to be implemented to Loviisa compared to the previously utilized partially low leaking one.In this paper, the changes in the reactor physical properties of the above mentioned different reloads are discussed. In the end, also an outlook to the possible future improvements to the fuel economy are given.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 20 10 – 14 October 2016, Helsinki, Finland

NEW ENGINEERING SAFETY FACTORS FOR LOVIISA NPP CORE CALCULATIONSKuopanportti, J., Saarinen, S., Lahtinen, T. and Ekström, K. Fortum Power and Heat Ltd, Espoo, Finland jaakko.kuopanportti@fortum.comABSTRACTIn Loviisa NPP, there are two limiting thermal margins called the enthalpy rise margin and the linear heat rate margin that are monitored during normal operation. The purpose of the first margin is to prevent bulk boiling and the second is meant to prevent fuel failures. Both of these margins are calculated with engineering safety factors. The factors take into account the effect of various manufacturing tolerances, impact of the irradiation and simulation uncertainties on the local heat rate and on the enthalpy of the coolant.The engineering factors were re-evaluated during 2015 and the factors were approved by the safety authority in 2016. The re-evaluation was performed by considering all of the identified phenomena that affect the local heat rate or the enthalpy of the coolant. These phenomena include: the calculation accuracy of the theoretical assembly power distribution and rod power distribution, the effect of various manufacturing tolerances of the assembly, the bowing of the assembly and the rods, the uncertainty of the coolant mixing within the assembly, the uncertainty of the coolant mass flow through an assembly, the uncertainties caused by the transformations of the fuel during irradiation, the uncertainty of the thermal power of the reactor.This paper summarizes the work that was performed during the re-evaluation of the engineering safety factors and presents the results for each uncertainty component. The new engineering safety factor of the linear heat rate is 1.115 when the old factor was 1.12. Whereas, the new engineering safety factor of the enthalpy rise margin is 1.100 when the old one was 1.16. The significant reduction of the engineering safety factor of the enthalpy rise margin will increase the fuel economy by about 1 %. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 21 10 – 14 October 2016, Helsinki, Finland

EXPERIENCES WITH 15M CYCLES IN PAKS NPPImre Nemes, István Pós, Tamás Parkó Paks NPP Ltd,HungaryABSTRACTUnits of Paks NPP are under transition from 12 to 15 mounts cycles in years 2015 and 16. Unit 2 and 3 started the 1st long cycles in 2015, Unit 1 and 4 in 2016. 4.7% enriched fuel containing 6 Gd pins was introduced to provide reactivity reserve for longer operation. During start-up measurements and normal operation reactor physical parameters was measured and compared to calculated ones. Assembly head coolant mixing model was checked as well as parameters of neutronic models. The presentation summarises and analyse all results concerning introduction of long cycles and 4.7% enriched fuel.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 22 10 – 14 October 2016, Helsinki, Finland

CORE LOADINGS OPTIMISATION IN DUKOVANY NPPJosef Bajgl, Martin Bárta CEZ Inc., Dukovany NPPABSTRACTThe core loading optimisation is an important step in the fuel cycle planning process. The continuous improvement of fuel cycle planning tools is demanded by changes in NPP operation conditions. There is described the fuel cycle planning process applied in Dukovany NPP in this paper. The attention is paid to improvement of core loading patterns optimisation program complex OPTIMAL/OPTIMAN especially. New program complex features and computational results for selected loading patterns are presented.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 23 10 – 14 October 2016, Helsinki, Finland

INFLUENCE OF TIME DURATION OF FAS IN SPENT FUEL POOL ON VVER-440 N-PH CHARACTERISTICS AFTER RELOADINGVáclav Krýsl, Pavel Mikoláš, Karel Vlachovský E-mail: pavel.mikolas@skoda-js.cz, Tel: +420 378042828 ŠKODA JS a.s., Orlík 266/15, 316 00 Plzeň, Czech RepublicABSTRACTPraxis (not only) at Dukovany NPP shows that some fuel assemblies can be loaded into core after some time duration in spent fuel pool. This is connected with a reload strategy – also with the fact that central FA is individual and if one assembly from six pieces has been used, the remaining five pieces cannot be reused simply on other positions in the core preserving usually requested core symmetry. But not only the central assembly is a question – sometimes it seems to be suitable to reload also more FAs from spent fuel pool, which were put off earlier (not only several weeks ago).The principle lies in the fact that quality (reactivity) of a (partially spent) FA being changed in time – this depends on measure of burn-up (also fuel enrichment) and time of placement in spent fuel pool. The reason is that concentrations of some isotopes (actinides and fission products) are being changed in time.Some isotope‘s changes (at least Xe135 and Sm149) must be considered in any case, but the others not so necessarily – in case of time of some weeks. But in case of longer time – they should be taken into account.Possible solution: To follow changes of concentrations of sufficiently big amount of isotopes (actinides and fission product) in producing code (so called “micro-depletion”). This approach was introduced on AER Symposium last year (DYN3D code). This is in any case a correct approach, but it is not easy to realize it (especially together with the pin-by-pin approach).The other possibility is to apply a suitable correction on FAs properties – change of reactivity (keff) due to time duration in spent fuel pool.Such approach has been applied in MOBY-DICK code and it is shortly described in the paper. Numerical results are also provided.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 24 10 – 14 October 2016, Helsinki, Finland

DEVELOPMENT OF FUEL CYCLES WITH NEW FUEL FOR VVER-440. FUEL RODS WITH 8.9 MM EXTERNAL DIAMETER: PRELIMINARY ASSESSMENT OF OPERATING EFFICIENCYА. Gagarinskiy, A. Lazarenko.NRC «Kurchatov institute», Moscow, RussiaABSTRACTEver since VVERs-440 have been introduced, their fuel assemblies are subject to ongoing improvements.Works intended to optimize the basic neutronic parameters of VVER-440 optional assemblies have started more than a decade ago.Until now, the basic structural parameters of fuel, such as rod diameter, have never changed (even in case of transition from the second to the third generation of fuel assemblies).The paper focuses on the calculated estimates of basic neutron-physical characteristics of the fuel cycle with fuel assemblies equipped with fuel rods which have reduced diameter up to 8.9 mm.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 25 10 – 14 October 2016, Helsinki, Finland

STATUS OF ALLEGRO PROJECT AT VUJEPetr DAŘÍLEK, Branislav HATALA, Radoslav ZAJAC VUJE Inc., Okruzna 5, SK 918 64 Trnava, Slovakia Petr.Darilek@vuje.skABSTRACTThe goal of the ALLEGRO Project is to design, build and operate the first Gas cooled Fast Reactor (GFR) Demonstrator. The GFR belongs to the 4th generation of nuclear reactors. Prototype of the gas cooled reactor ALLEGRO is a subject of common interest of association of countries France, Slovakia, the Czech Republic and Hungary with the goal to deploy prototype of the reactor in Central Europe and present its functionality.In this article are aimed several ALLEGRO core modifications focused on criticality feasibility related to 20 % UOX fuel. Several cases for the criticality analyses are defined. These cases differ in the following parameters – number of fuel rings and fuel height.The last part of the article is proposed to technical description of the first Slovakian helium loop and experimental programme performed at this loop. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 26 10 – 14 October 2016, Helsinki, Finland

AER WORKING GROUP C ACTIVITIES IN 2016Imre Nemes Paks NPP Ltd nemes@npp.huABSTRACT Review of AER Working Group C Meeting is given.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 27 10 – 14 October 2016, Helsinki, Finland

16 YEARS OF RELIABLE OPERATION OF THE SCORPIO-VVER CORE MONITORING SYSTEM AT SLOVAK AND CZECH NPPsJozef MolnárÚJV Řež, a. s., Hlavní 130, Řež, 250 68 Husinec, Czech Republic jozef.molnar@ujv.czABSTRACTThe SCORPIO-VVER core monitoring and surveillance system is a software based system implemented on the robust mission critical hardware platform connected to the plant data highway. The main task of the system is the continuous core monitoring and core characteristics evaluation to help safely and effectively operate the nuclear reactor.The system was developed by local organizations as a full replacement of the original Russian VK3 system with the goal to strengthen the reactor’s core monitoring and surveillance. Since 1999 the SCORPIO-VVER system is a valuable tool for the reactor operators and reactor physicists and was licensed by both Czech and Slovak Nuclear Regulatory Bodies as a Plant Technical Specification Surveillance Tool.The development of SCORPIO-VVER core monitoring system continues along with the changes in VVER reactors operation. Within the planned upgrades the system is being adapted according the utility needs. Between the most significant upgrades belongs the modifications in connection with implementation of a new digital I&C system, adaptation of the system to up-rated unit conditions, loading new generation fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), in design and methodology of the limit and technical specifications checking and improvements in the predictive part of the system.Since the first installation the SCORPIO-VVER system has a remarkable operating history and experience. More than 16 years of experiences on 6 units of VVER-440 type of reactors, at two different NPPs, in two different countries helps the SCORPIO-VVER Developer’s Team to put the system to a very high level of quality and reliability.Even if the system is installed only on VVER-440 reactors, it could be adapted to the needs of other VVER type of reactors (VVER-1000) and to needs of education and training centers too. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 28 10 – 14 October 2016, Helsinki, Finland

VERIFICATION AND DEVELOPMENT EXPERIENCE OF CORE MONITORING SYSTEM (CMS) PHYSICAL CALCULATION CODES FOR WATER COOLED WATER MODERATED REACTORS AT THE SOFTWARE AND HARDWARE INSTRUMENT “KRUIZ” BASISKuzhil A.S., Padun S.P., Bykov A.V. Innovative firm SNIIP Atom LLCABSTRACTSoftware and hardware instrument “Kruiz” was developed by our firm at the end of 1990-s. Their last versions are currently implemented at Russian, Ukrainian and EU (Mochovce) NPP. Over time the great scope of verification and validation of core monitoring system (CMS) software was done. Due to the new regulation requirements certification of calculating kernel of CMS “Kruiz” software in Russian nuclear regulator is held. Verification peculiarities of some kernel calculation components are presented at the report. Comparison examples with certified codes and Russian and abroad NPP operation measured data are presented. Ability of further software and hardware instrument “Kruiz” development for the PWR core monitoring was shown. To date our “Kruiz” CMS software has reference operation experience on WWER reactors more than 250 reactor-years.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 29 10 – 14 October 2016, Helsinki, Finland

MODELS OF PIN-WISE RECONSTRUCTIONS IN VERONA SYSTEM AND ITS VALIDATION BY MIDI-CORE BENCHMARKIstván Pós, Zoltán Kálya, Tamás Parkó and Sándor Patai-Szabó MVM Paks NPPABSTRACTThis year the installation of the modernized VERONA core monitoring system (version V7.0) will be finished at the NPP Paks. The VERONA applies several new models for calculations of neutron flux and thermal power inside the reactor core. One of the most important models is the pin-wise reconstruction which plays very important role both in VERONA and in off-line core calculations.The modernized pin-wise model can take into account larger area of assemblies and applies more precise spatial resolution during its calculations, than it was used in the earlier model. To describe the effect of gadolinium caused by its big neutron absorption we applied pin-wise discontinuity factors calculated by HELIOS code. The accuracy of the pin-wise reconstruction model has been increased in great extent by utilization of node-wise fluxes as boundary conditions at the border of the considered regions of assemblies as well.During the process of validation several comparisons have been done. The pin-wise power distributions calculated by new model were compared to references given by MCNP or HELIOS codes. This paper gives detailed information about the results of MIDI-core benchmark where the renewed pin-wise reconstruction model was compared to MCNP results.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 30 10 – 14 October 2016, Helsinki, Finland

INVESTIGATION OF CIRCULATING TEMPERATURE FLUCTUATIONS OF THE PRIMARY COOLANT IN ORDER TO DEVELOP AN ENHANCED MTC ESTIMATOR FOR VVER-440 REACTORSSándor KISS, Sándor LipcseiCentre for Energy Research, Hungarian Academy of Sciences P. O. Box 49, H-1525, Budapest, Hungary sandor.kiss@energia.mta.hu, sandor.lipcsei@energia.mta.huABSTRACTOur aim was to develop a method applicable to use in normal operation and based on noise diagnostics for estimation of the moderator temperature coefficient of the reactivity (MTC) for the Paks VVER-440 units. The method requires determining core averaged neutron flux and temperature fluctuations. Circulation period of the primary coolant, transfer properties of the steam generators, as well as the source and the propagation of the temperature perturbations and the proportions of the perturbation components were investigated in order to estimate the feedback caused by the circulation of the primary coolant. Finally, after developing the new MTC estimator, determining its frequency range and optimal parameters, trends were produced based on an overall evaluation of measurements made with standard instrumentation during a one year long period of all four units of the Paks NPP.Keywords: MTC estimation, VVER-440, circulation period of the coolant, propagating perturbations, transfer properties of the steam generator, average loop model26th Symposium of AER on VVER Reactor Physics and Reactor Safety 31 10 – 14 October 2016, Helsinki, Finland

THE VERIFICATION RESULTS OF METHODOLOGY FOR DETERMINING THE WEIGHTED MEAN COOLANT TEMPERATURE IN THE PRIMARY CIRCUIT HOT LEGS OF WWER-1000 REACTOR PLANTSIu. Saunin, A. Dobrotvorski, A. Semenikhin, A. Korolev, S. Ryasny JSC “ATOMTECHENERGO”, Mytishchi, RussiaABSTRACTThe new methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants has been developed by experts of JSC “Atomtechenergo”. It was represented at 25th Symposium АER on WWER reactor physics and safety research. The new methodology development was defined by necessity to decrease error in calculation of the weighted mean coolant temperature in hot legs due to the coolant temperature stratification. The executed experimental and calculating researches results have been considered in the developed methodology.Methodology verification has been fulfilled by comparison of calculations results received with using and without using methodology in various operational conditions of three WWER-1000 power units. These power units have differences in structure and placing of temperature detectors and also differed in fuel loading. For the aims of verification 39 operational conditions have been analysed. They were differed in reactor power levels, positions of control rod regulating group and etc.The received results of verification have confirmed that with using of new methodology the objective decreasing of error in determining the weighted mean coolant temperature in the primary circuit hot legs is reached. The amount of decreasing depends on the character of stratification which is various for different objects. The maximum decreasing of error among investigated objects during verification was 0,6 °С.The executed verification shows that it is necessary to continue researches of coolant temperature stratification in hot legs and possibilities of new methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 32 10 – 14 October 2016, Helsinki, Finland

CONSIDERATION OF INFLUENCE OF THE GAP SIZE (BETWEEN FUEL ASSEMBLIES) CHANGES ONTO POWER FIELD DISTRIBUTION IN VVER-1000 REACTORS WITH THE HELP OF PERMAK-A CODES. Aleshin, M. Kalugin, A. Pinegin, A. Ryzhov NRC “Kurchatov Institute”Moscow, RussiaABSTRACTThe gap sizes between FAs in VVER-1000 reactors can diverge from its design values due to influence of some factors. The changes in gaps between fuel assemblies lead to the changes in uranium-water ratio in some areas of fuel rods lattice. Hence, they lead to the changes in power of pins that located in peripheral rows of FA. In the paper it is presented methods and results of pin-by-pin power distribution numeric calculations using diffusion code PERMAK-A. The results of calculations using presented methods are well-corresponded with the results of precision calculations by Monte Carlo program complex MCU for the mini-core with seven FAs. Obtained results allow to carry out the series of pin-by-pin power distribution calculations with taking into account possible technological changes in gaps between FAs in the core with the help of PERMAK-A code. In the paper it is presented the results of test calculations for the core.The procedure of taking into account the changes in pin power due to changes gaps between FAs is included into STAT-TVS code. STAT-TVS code is intended for estimation (using Monte Carlo methods) the errors in power distribution, the minimum values of DNBR and maximal values of fuel temperatures in FA accounting technological and calculation errors. In the paper it is presented the results of statistical simulation neutron-physical and thermo- physical processes in FA. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 33 10 – 14 October 2016, Helsinki, Finland

SHUTDOWN MARGIN CALCULATIONS FOR THE DETERMINATION OF THE LIMITING POSITION OF CONTROL FUEL ASSEMBLIES ON DUKOVANY NPPMartin Šašek, Zdeňka Neduchalová, Svatobor Štech E-mail: martin.sasek@skoda-js.cz, Tel: +420 378 042 835 ŠKODA JS a.s., Orlík 266/15, 316 00 Plzeň, Czech RepublicABSTRACTNew realistic approach for the determination of control fuel assembly (CFA) position limiting curve presented last year on AER symposium resulted in lower CFA positions than previous conservative approach. However, further investigation of Dukovany NPP operations showed that a limiting curve based on shutdown margin (SDM) and CFA ejection could lead to more flexibility for the NPP operators during transition processes (TP) without compromising the NPP’s operation safety. That’s why comprehensive SDM calculations were performed for static calculations and transition processes of the realistic approach. Coarse mesh calculations were used for basic orientation in the problem and were followed by fine mesh calculations of selected scenarios. Results of these fine mesh calculations and their analysis are presented in this contribution. Analysis is aimed mainly on finding out which scenarios and at which conditions are the limiting ones. Considered conditions include time in cycle at which the TP began, TP power or elapsed time during the TP. As a conclusion, new limiting curve based on the shutdown margin calculation is proposed together with information on how this new approach could be included in the current operation procedures of the Dukovany NPP. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 34 10 – 14 October 2016, Helsinki, Finland

AER WORKING GROUP D ON VVER SAFETY ANALYSIS – REPORT OF THE 2016 MEETINGS. Kliem Helmholtz-Zentrum Dresden-Rossendorf Institute of Resource EcologyP.O.B. 51 01 19, D-01314 Dresden, Germany S.Kliem@hzdr.deABSTRACTThe AER Working Group D on VVER reactor safety analysis held its 25th meeting in Villigen, Switzerland, during the period 30-31 May, 2016. The meeting was hosted by PSI Villigen and was held in conjunction with the 10th workshop on the OECD Benchmark for Uncertainty Analysis in Best-Estimate Modelling (UAM) for Design, Operation and Safety Analysis of LWRs. Altogether 19 participants attended the meeting of the working group D, 9 from AER member organizations and 5 guests from non-member organization. The co-ordinator of the working group, Mr. S. Kliem, served as chairman of the meeting.The meeting started with a general information exchange about the recent activities in the participating organizations.The given presentations and the discussions can be attributed to the following topics: Safety analyses methods and results Code development and benchmarking Thermal hydraulic analyses of passive safety systems Future activitiesA 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 chairman.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 35 10 – 14 October 2016, Helsinki, Finland

THE PROCESS OF ASYMMETRIC BORON DILUTION AT VVER-1000 DURING THE QUICK BORON INJECTION SYSYTEM OF KUDANKULAM NPP. EXPERIMENTAL RESULTS AND SIMULATION.S.V. Tsyganov, A.V. Kotsarev, A.V. Baykov NRC “Kurchatov Institue”ABSTRACTDesign of Kudankulam NPP units includes additional and unique for VVER Quick Boron Injection System (QBIS) for beyond design-basis accident management without scram. During the physical start-up stage at hot zero power of both Kudankulam units special tests were performed to assess the efficiency of the system. In the course of test three out of four QBIS tanks had been promptly opened and it lead to asymmetrical injection of boric acid into the core.The simulation of the test process, including ex-core ion chambers currents and reactivity, has been accomplished using ATHLET/BIPR-VVER code. Behaviour of some reactor parameters in the course of test and some results of simulation are discussed in the presentation.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 36 10 – 14 October 2016, Helsinki, Finland

BEPU APPROACH FOR WWER SAFETY ANALYSIS IN RIAI.Bilodid, I.Ovdiienko, M.IeremenkoState Scientific and Technical Centre on Nuclear and Radiation Safety, Kyiv, Ukraine Stusa st. 35-37, 03142 Kyiv, Ukraine yn_ovdienko@sstc.com.uaT. LötschTÜV SÜD Industrie Service GmbH, Energy and Systems (IS-ES), Munich, Germany Westendst. 199, 80686 Munich, Germany thomas.loetsch@tuev-sued.deABSTRACTAt present time, Ukraine faces the problem of small margins of acceptance criteria in connection with the implementation of a conservative approach for safety evaluations. The problem is particularly topical conducting feasibility analysis of power up-rating for Ukrainian nuclear power plants. Such situation requires the implementation of a best-estimate approach on the basis of an uncertainty analysis. For some kind of accidents, such as loss-of-coolant accident (LOCA), the best estimate approach is, more or less, developed and established. However, for reactivity initiated accident (RIA) analysis an application of best estimate method could be problematical. A regulatory document in Ukraine defines a nomenclature of neutronics calculations and so called “generic safety parameters” which should be used as boundary conditions for all VVER-1000 (V-320) reactors in RIA analysis. In this paper the approach of realization of generic safety parameters as uncertainty parameters for RIA analysis with the use of the stand-alone 3D neutron kinetic code DYN3D, coupled version of DYN3D/ATHLET and the GRS SUSA approach are presented.The results of mentioned approach application in comparison a conservative approach results are presented.The work was performed in the framework of the project BMUB/GRS 3614R01520 under German BMU support. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 37 10 – 14 October 2016, Helsinki, Finland

THE SEVENTH AER BENCHMARK CALCULATION WITH HEXTRAN-SMABRE-PORFLOAnitta Hämäläinen, Ville Hovi, Elina Syrjälahti, Hanna Räty, VTT Technical Research Centre of FinlandABSTRACTThe seventh AER benchmark is the first AER Working group D benchmark, where three dimensional thermal hydraulic codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. First the 7th BM is calculated with the internally coupled HEXTRAN-SMABRE reactor dynamics code. In the next stage one-way coupled analyses with the Porous CFD code PORFLO is performed. The results of these analyses are applied in the second HEXTRAN- SMABRE analyses, reported here. These three codes are developed at VTT.The 7TH BM consists of start-up of the sixth, isolated loop in VVER-440 plant. The isolated loop contains initially cold water without boric acid and start-up leads to somewhat asymmetrical core power increase due to feedbacks in the core. The analyses model is basically the VVER-440 plant model used for the 5th and 6th BMs at VTT. The modelling issues for this BM are reported and some evaluation against the earlier reported comparisons between the system codes is done.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 38 10 – 14 October 2016, Helsinki, Finland

THE PROCEDURES FOR SEARCHING CONSERVATIVE INITIAL CORE STATES FOR SOME REACTIVITY INITIATED ACCIDENTS (RIA) TAKING INTO ACCOUNT XENON TRANSIENTSA.A. Pinegin, S.V. SemenovNational Research Centre “Kurchatov institute” Moscow, RussiaABSTRACTTaking into account energy field redistribution upon xenon transients is important problem of fuel cycles design and their safety analysis. Under fuel cycle design it is provided considerable reserves from limit values of local energy field. These reserves are intended to provide safety passing of xenon transients.In practice there are problems with accuracy of xenon transients simulating, because xenon transients have extreme sensitivity to some physical characteristics. Particularly, temperature reactivity coefficients and absorption cross-section of xenon-135 belong to such characteristics. The changes in these characteristics can lead to significantly changes in the course of xenon transients simulating. In the report it is discussed problems of accounting observable errors in xenon transients simulating during searching conservative initial states for some reactivity accidents.Reactivity accident consequences can significantly differ depending on scenario of xenon transient, in the course of which it was taken place a violence of equipment operation. In the report it is discussed different possible statistics procedures of searching conservative scenarios of xenon transients. For simplicity of analysis it is supposed that scenarios of xenon transients are determined by several parameters (initial power level, unloading power depth, initial position of CR CPS and etc.)In the report it considers the procedure of searching of conservative scenarios of xenon transients using statistical simulation of set of parameters, which determine the course of xenon transients, and initiation of violence of equipment operation within different time moments. Such approach allows to obtain effective statistical estimation of conservative scenarios characteristics. To searching global conservative scenarios of xenon transients it is analyzed the possibility of using simulated annealing algorithm. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 39 10 – 14 October 2016, Helsinki, Finland

OPTIMIZATION OF MICROSCOPIC DEPLETION METHODOLOGY IN DYN3D CODE FOR REACTIVITY CALCULATIONSY. Bilodid1, D. Kotlyar21Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden, Germany Tel.: +49 351 260 2020Fax: +49 351 260 3299 y.bilodid@hzdr.de2Department of EngineeringUniversity of Cambridge, CB2 1PZ Cambridge, United KingdomABSTRACTNodal diffusion codes such as DYN3D obtain homogenized few-group macroscopic reaction cross sections (XS) of coarse-mesh space elements (nodes) from XS-libraries, which are generated using lattice neutron transport codes. Typically, fuel depletion is simulated by a lattice code in 2D single assembly model using core average operating conditions (moderator density, fuel temperature, control rod presence etc.).However, the local operating conditions in the core nodes may differ significantly from the core average values. XS generated using a depletion calculation under core averaged conditions neglect the local variations of the spectrum history and should be corrected. In order to account for the local spectrum history effects a number of methods, including microscopic depletion and various formulations of spectral indexes was developed and implemented in nodal codes.Recently a new hybrid method was developed and implemented in DYN3D, which combines the generalised micro-depletion correction with Pu-history indicator. The detailed nuclide content (over 1000 nuclides) is calculated by DYN3D and utilized to correct macroscopic XS, while Pu-correction is applied to the isotopic microscopic cross sections. This general approach not only accurately accounts for fuel spectral history but also provides detailed isotopic content for spent fuel inventory, radiotoxicity and decay heat applications. On the other hand, tracking of the detailed nuclide content is computationally expensive in full core burnup simulation and excessive for spectral history accounting.This paper describes an optimisation of a hybrid micro-depletion method for practical reactor simulations. The nuclide inventory estimated for each node is limited to nuclides which influence fuel reactivity. The transmutation matrix solved by DYN3D was simplified to about 200 nuclides using HELIOS 2 isotopic cross section library. HELIOS 2 was also used to generate homogenised macro- and microscopic cross section for DYN3D in demonstrated test cases. The optimised hybrid micro-depletion method was verified on various spectral history effects against HELIOS 2 reference.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 40 10 – 14 October 2016, Helsinki, Finland

POSSIBLE BEYOND DESIGN-BASIC ACCIDENT ANALYSIS DEALING WITH CRIPPLING OF VVER-1200 FUEL ASSEMBLIESSIKORIN S., POLAZAU S., DAMARAD Yu. Joint Institute for Power and Nuclear Research – Sosny,National Academy of Science of BelarusABSTRACTThe destruction variants for TVS-2M fuel assembly of reactor VVER-1200 are analyzed on the possible beyond design-basic accident at a handling and transportation of fuel assemblies, including unauthorized action. By means of program codes MCU-PD and MCNP4C the analytical researches were executed on criticality of multiple systems on the basis of fuel rods, contained in fuel assemblies. The opportunity to create nuclear-dangerous system with Keff exceeding 1,09 in the water are shown. Technical decisions for avoidance of super-critical system formations and fulfillment of nuclear safety requirements are offered.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 41 10 – 14 October 2016, Helsinki, Finland

MULTISCALE APPLICATION OF SYSTEM CODE ATHLET FOR DETAILED VVER CORE ANALYSISS. Nikonov1, I. Pasichnyk2, K. Velkov21GRS scientific guest, Moscow, Russia2Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching, GermanyABSTRACTThis paper sums up intermediate results of research efforts aimed to develop a detailed in-vessel ATHLET VVER-type reactor model. It is shown, that together with the development of the input data deck preprocessor, this activity also requires substantial changes in the ATHLET code itself with a goal to accelerate calculations and reduce wallclock time of the simulations. The results of the code evolution are reported and it is shown that for the generic VVER input data deck the speedup factor of 170 is achieved. The turbulent model which takes into account turbulent mass exchange between connected thermohydraulic objects with different level of nodalization is tested. A testbed for the model development and verification serves the OECD/NEA benchmark Kalinin-3 (K3). Calculations of the coolant state distributions in the active core are carried out for assemlbywise as well as pinwise nodalization of the selected fuel assemblies. An achieved progress allows within one system code ATHLET to cover wide spectrum of problems starting from the detailed description of processes in the fuel assembly head up to the description of global parameters of primary and secondary loops of the nuclear reactor facility. It substantially increases the area of possible ATHLET code applications compared to its initial purposes defined by the first designers of the code. Future directions are outlined which mainly concerns the development of the simulation ecosystem which will be able to perform coupled neutronphyiscal and thermohydraulic pin-by-pin calculations.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 42 10 – 14 October 2016, Helsinki, Finland

A METHODOLOGY FOR THE ESTIMATION OF THE RADIOLOGICAL CONSEQUENCES OF A LOCA EVENTAndrás Keresztúri, Ádám Brolly, István Panka, Tamás Pázmándi, István Trosztel MTA EK, Centre for Energy Research, Hungarian Academy of Sciences H-1525 Budapest 114, P.O. Box 49, Hungary andras.kereszturi@energia.mta.huABSTRACTFor calculation of a the radiological consequences of LOCA events a set of various computer codes modeling the corresponding physical processes, disciplines and their appropriate subsequent connections are necessary. For demonstrating the methodology applied in MTA EK, the LOCA event at a shut down reactor – when only limited configuration of ECCS is available – was selected. In this special case, partial fission gas release from a number of fuel pins is obtained from the analyses. The presentation overviews the the initiating event, the corresponding thermal hydraulic calculations, the physical processes, the necessary models and computer codes, their connections, the applied conservative assumptions and the B+U evaluation applied for characterizingthe pin power and burnup distribution in the core, the fuel behavior processes, the calculations applied to predict whether the fuel pins aregetting in-hermetic, the results of the activity transport and dose calculations.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 43 10 – 14 October 2016, Helsinki, Finland

MONTE CARLO CALCULATION OF 24NA, 42K AND 16N ISOTOPES IN THE COOLANT OF VVER-440 NUCLEAR REACTORSG. Radócz, A. Gerényi, I. SzalókiInstitute of Nuclear Techniques, Budapest University of Technology and Economics, Budapest, HungaryABSTRACTThe leaking condition of steam generators is a very important safety parameter for the commissioning of pressurized water reactors. As part of the monitoring of the operating parameters of steam generators the detection of leaking state of heat transfer tubes. The leaking state is crucial to avoid the decontamination of the secondary water circuit; therefore different monitoring systems are available to ensure the detection of the leakage of the primary coolant. Gross gamma detection systems of 16N isotope were installed by the beginning of the main steam collectors at the secondary loops. The main role of this measuring arrangement is to give warning signal when the calculated concentration of the 16N in the secondary loops reaches the limit of allowed leakage level.The revision of restrictions rules of the primary-secondary leakage in steam generators was performed in the framework of a R&D project in Paks NPP. The specific activity of the relevant isotopes (16N, 24Na, 42K) produced by neutron induced nuclear reactions in the primary coolant were determined by Monte Carlo burnup calculations for different reactor cores built from various enrichment fuel assemblies. The activities in the secondary loops caused by the primary-secondary leakage were calculated by dispersion model during normal operation. The activities calculated by MCNP6 simulation code were compared with the results of radiochemical measurements on coolant samples originated from the primary and the secondary loops to verify our model calculations. According to this comparative analysis the agreement between the measured and calculated specific activities was within the statistical errors.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 44 10 – 14 October 2016, Helsinki, Finland

RECALCULATING THE STEADY STATE CONDITIONS OF THE V-1000 ZERO-POWER FACILITY AT KURCHATOV INSTITUTE USING MONTE CARLO AND NODAL DIFFUSIONVille SahlbergVTT Technical Research Centre of FinlandABSTRACTContinuous-energy Monte Carlo reactor physics code Serpent 2 was used to model the steady state conditions measured in V-1000 zero-power critical facility at Kurchatov Institute, Moscow, in 1990-1992. The Serpent 2 results were compared to the steady state power measurements of the test facility with all control rods withdrawn from the core. Serpent 2 was used to generate group constants for reactor dynamics code HEXTRAN and a HEXTRAN calculation of the test facility was performed.The original measurements were carried out by inserting partial-length fuel rods into the core. The Serpent 2 calculation allowed direct calculation of the relative powers of these rods instead of relying on pin power reconstruction. The properties of the radial reflector and the gaps between the radial reflector and the fuel assemblies of the test facility were examined with additional detail. The relative power distribution of the zero-power steady state was found to be very sensitive to the properties of the radial reflector. Nodal diffusion calculations carried out previously by several different groups in 2003 found a significant power tilt in the core. This was reproduced in the Monte Carlo calculation. However, the reflector gaps produced another power tilt in addition to the previously known one.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 45 10 – 14 October 2016, Helsinki, Finland

DIFFERENTIAL REACTIVITY COEFFICIENTS ESTIMATION METHOD FOR VVER REACTOR BY KORSAR/GP AND TRAP-KS CODES APPLICATIONM.A. Uvakin, A.P. Demekhin, G.V. Alekhin, V.V. Bryukhin, A.N. Ustinov JSC OKB GidropressABSTRACTThis work is oriented to actual problem resolving. The problem is VVER reactor facility stationary conditions preparing for coupling physical and thermo-hydraulic dynamical processes calculation. Work contains fuel temperature and coolant density reactivity coefficients estimation approach. It is based on finite-difference approximation. Developed model is built on macroscopic library structure and feedback parameterization principles. Method is suitable for KORSAR/GP and TRAP-KS codes which are used physical data preparing by SAPFIR_95 code. Material contains some application results of introduced method for stationary conditions and for transients with power shifting.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 46 10 – 14 October 2016, Helsinki, Finland

CFD ANALYSES OF THE ROD BOWING EFFECT ON THE SUBCHANNEL OUTLET TEMPERATURE DISTRIBUTIONKaroliina Ekström. Timo ToppilaFortum Power and Heat Ltd, FinlandABSTRACTIn the Loviisa 1 and 2 nuclear power plants, the subcooling margin of the hottest subchannel is monitored during normal operation. The temperature of the coolant in the hottest subchannel of the fuel assembly is limited to the saturation temperature. In this work the effect of bent fuel rods to temperature of the coolant in the hottest subchannel is studied.Bending of the rods occurs during normal operation due to differences in the heat profiles in the rods. Due to the bending of the rods the flow area of the subchannels varies. The coolant temperature will rise more in the subchannel with smaller flow area and this has to be taken into account in the safety margin of subchannel enthalpy rise. The maximum bow of rods in a normal fuel assembly in Loviisa NPP has been estimated based on visual observations. Computational Fluid Dynamics (CFD) simulations are used to estimate how much the estimated maximum bow of a rod affects the temperature rise of the subchannel.The quantitative uncertainty of the predicted enthalpy rise in fuel bundle subchannel is estimated based on the uncertainty of modelling of mixing between subchannels. The measured turbulence quantities from LDA measurements of cold test assembly made in 1990s in Fortum are compared with CFD results to give uncertainty estimation for turbulence, which is further used for quantitative uncertainty estimation of mixing and simulated subchannel enthalpy rise.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 47 10 – 14 October 2016, Helsinki, Finland

EXPERIMENTAL AND NUMERICAL THERMAL-HYDRAULICS INVESTIGATION OF A MOLTEN SALT REACTOR CONCEPT COREBogdán Yamaji, Attila AszódiInstitute of Nuclear Techniques, Budapest University of Technology and Economics tel: +36 1 463 2112 fax: +36 1 463 1954e-mail: yamaji@reak.bme.hu, aszodi@reak.bme.huBudapest, Műegyetem rkp. 9., Hungary 1111ABSTRACTIn the paper measurement results of experimental modelling of a molten salt fast reactor concept will be presented and compared with three-dimensional computational fluid dynamics (CFD) simulation results. The investigated molten salt reactor concept has a homogeneous cylindrical core without any internal structures. Previous measurements and simulation results have shown that this core geometry could be optimized for better thermal-hydraulics characteristics. In case of the original geometry strong unfavourable flow separation could develop. Two regions would form: a slow, stagnating region near the cylindrical wall due to recirculation and a high velocity jet stream. Such a disadvantageous velocity distribution would also lead to unfavourable temperature distribution in the core and it could negatively affect the characteristics of the core from neutronics point of view as well.The experimental investigation was carried out on the scaled and segmented plexiglas model of the molten salt reactor concept. In the scaled and segmented mock-up the working fluid is water. An internal flow distributor plate was designed and installed with the purpose of optimizing the flow field in the core by enhancing its uniformity. Measurement results of the experimental model will be presented and compared to numerical simulation results with the purpose of validating the CFD models. For the non-intrusive flow measurements carried out on the scaled and segmented mock-up particle image velocimetry (PIV) method was applied.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 48 10 – 14 October 2016, Helsinki, Finland

CFD ANALYSIS ON THE TEST SECTION OF SCWR-FQT DEMO LOOPBence Mervay, Attila KissBudapest University of Technology and Economics, Institute of Nuclear Techniques, 1111 Budapest, Muegyetem rkp. 9. R.317/7a, kissa@reak.bme.huABSTRACTThe CFD analysis presented in this paper is connected to the SCWR-FQT project funded by the EU that was finished in 2014 and aimed at the design and the licensing of a test loop that demonstrates the viability of the fuel bundle model of the European SCWR in real (radioactive) environment. The goal of the demonstration was to prove the compliance of the material of the fuel rod’s cladding (whether it is able to withstand the chemically aggressive, high pressure and temperature supercritical water) and the thermohydraulic adequacy of the fuel bundle geometry with helical spacers. The BME NTI participated in the thermohydraulic design of the test rod bundle. One of the Institute’s tasks was to participate in a blind CFD benchmark for which the measurement results were provided by one of the Chinese partners. The measurements were conducted in the previously built SWAMUP test loop in which the test section was placed, that was actually the SCWR-FQT fuel bundle geometry, scaled up by a factor of 1.25. Due to the relatively complex geometry the numerical grid is mainly unstructured tetrahedral grid. The heat conduction was modelled in the solid domains. A mesh sensitivity study was conducted as well as a turbulence model sensitivity study and a boundary layer sensitivity study.After the completion of the project and thus the benchmark the Institute carried on dealing with the problem, we tried to improve our model. We also examined different geometries that differed in the number of revolution of the helical spacers. In this way we could examine the thermohydraulic differences between the geometries with different number of revolution. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 49 10 – 14 October 2016, Helsinki, Finland

POOLSIDE INSPECTIONS OF SPENT NUCLEAR FUEL AT LOVIISA NPPIiro-Ville Lehtinen Fortum Power and Heat OyABSTRACTLoviisa NPP has a long history of poolside inspections of spent nuclear fuel starting from the middle of the 80’s. These inspections are usually performed at the spent nuclear fuel interim storage building after one to two years cooling period. The inspection equipment can be moved to the reactor building well also, if it is required to perform inspections right after the removal of the fuel from the core. Loviisa NPP uses own inspection equipment originally developed by IVO Engineering (now part of Fortum).The results of the poolside inspections are used to verify fuel performance in normal operating conditions and during interim storage, but also to evaluate the effects of fuel design changes and to find root causes of fuel failures. Furthermore, the results have been used together with reactor physics and fuel performance codes to improve the fuel economy and to verify safety analysis input parameters.In this paper Loviisa NPP poolside inspection equipment and methods are introduced together with a short summary of the current inspection programs. A summary of inspection results is also given. In addition, some examples are presented on how the poolside inspection results have helped Loviisa NPP to better understand fuel performance, reduce fuel failures and improve fuel economy.Keywords: VVER-440, fuel performance, operational experience, poolside inspections26th Symposium of AER on VVER Reactor Physics and Reactor Safety 50 10 – 14 October 2016, Helsinki, Finland

FUEL BEHAVIOR CALCULATIONS FOR DEVELOPED FUEL AND CYCLE AT PAKS NPP, HUNGARYGábor BÓNA, Imre NEMES, Botond BELICZAI Paks NPP, Hungary bonag@npp.huABSTRACTFrom 2015 15 months cycles are being introduced at Paks NPP. For this, the 4.7% enrichment fuel assembly (with 6 Gd pins) is used. For these long cycles the economic solutions can only be improved by increasing the water – Uranium ratio. New fuel geometry is proposed with some modifications such as smaller clad thickness and outer clad diameter, the lack of the central hole and higher average enrichment. For the new assemblies the loading patters are determined and the required neutronic and thermo-hydraulic calculations were made. In my presentation I would like to show the results of the fuel behavior calculations, which were carried out for this new fuel type. We found that the main points are the identification of the pins with the highest parameters (fuel temperatures, dimension changes, fission gas release and hoop stress) and the question of the start-up speed. We developed a tool, which allows us to calculate the fuel behavior parameters for all assemblies and all pins in the core. The results are very interesting. The identification of the most “loaded” pins caused some surprises, which justifies the need for the full core fuel behavior calculations. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 51 10 – 14 October 2016, Helsinki, Finland

CURRENT STATE OF FUEL PERFORMANCE MODELLING IN THE LOVIISA NPPVille PeriFortum Power and Heat OyABSTRACTSafe operation of nuclear fuel in the Loviisa nuclear power plant is ensured not only through years of operational experience and development, but also through understanding fuel behavior in normal operating conditions, transients and accidents. In the Loviisa nuclear power plant, steady state and transient fuel behavior is currently predicted and analyzed with the TRANSURANUS fuel performance code. TRANSURANUS has been used in fuel performance calculations on Loviisa fuel rods since the year 2014. The initial analyses focused on reproducing and re-evaluating prior results from ENIGMA fuel performance calculations at Fortum and VTT, in addition to comparing the results of the TRANSURANUS calculations to data from poolside and hot cell investigations.Further and current analyses concentrate on three different subjects. Firstly, TRANSURANUS is used in justifying the safety of fuel operation in all normal operating scenarios. Even in the most limiting cases, all parameters of fuel behavior are within their respective limits. Secondly, best-estimate analyses with TRANSURANUS provide reference results for poolside investigations, where calculated fuel parameters are used in determining fission gas release ratios from gamma scanning data and as a comparison to the measured fuel properties. Thirdly, results of fuel performance analysis serve as input data for neutron-physical and process simulation codes. This paper includes a short introduction on fuel performance analyses at Fortum with a summary of the calculation process. The principles and main results of the three primary areas of analysis are reviewed. The paper is concluded with an outlook on future analyses.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 52 10 – 14 October 2016, Helsinki, Finland

AER WORKING GROUP E ACTIVITIES IN 2016Radoslav ZAJAC, Vladimír CHRAPČIAK VUJE Inc., Okruzna 5, SK 918 64 Trnava, Slovakia Radoslav.Zajac@vuje.skABSTRACTReview of AER Working Group B Meeting in UJV Řež, Czech Republic is given. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 53 10 – 14 October 2016, Helsinki, Finland

ASSESSMENT OF THE REACTIVITY BIAS AND BIAS UNCERTAINTY DUE TO WWER-440 FUEL DEPLETION UNCERTAINTIESN. Baghdasaryan, S. Bznuni, A. AmirjanyanNuclear and Radiation Safety Center,P. KohutBrookhaven National LaboratoryJ. RamsayUSA Nuclear Regulatory CommissionABSTRACTIn this paper, bias and bias uncertainty in the neutron multiplication factor due to biases and bias uncertainties in the calculated nuclide concentrations in the spent nuclear fuel of WWER- 440 type was assessed. To determine isotopic biases and bias uncertainties in the calculated nuclide concentrations, they were compared to the results of the measurements of isotopic compositions from destructive radiochemical assay of WWER-440 spent fuel carried out in the RIAR (Dimitrovgrad). By employing Monte Carlo uncertainty sampling method, the isotopic biases and bias uncertainties were applied to the spent fuel compositions of ANPP spent fuel transport cask model to determine bias and bias uncertainty values in keff.The MCNP 6.1 code and the ENDF/B-VII.1 nuclear data were used to assess keff bias and bias uncertainty results for ANPP spent fuel transport cask model.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 54 10 – 14 October 2016, Helsinki, Finland

SIMULTANEOUS APPLICATION OF X-RAY FLUORESCENCE AND GAMMA SPECTROMETER FOR ANALYSIS OF RADIOACTIVE WASTE MATERIALAnita Gerényi, Gábor Radócz, Imre SzalókiInstitute of Nuclear Techniques, Budapest University of Technology and Economics, 1111 Budapest, HungaryABSTRACTDetermination of the activity of the radioactive isotopes and concentration of inactive elementary composition in waste materials should be the basis of the identification for future management. For this purpose a new combined mobile X-ray fluorescence spectrometer (XRF) with silicon drift detector (SD) and gamma spectrometer using cadmium-zinc-telluride (CZT) detector was developed. The XRF analysis provides the chemical composition of the sample matrix. For calculation of the quantitative composition a new physical model based on the adaptation of the fundamental parameter method (FPM) was developed. For the numerical calculations new software was designed in MATLAB operation environment. This analytical method is capable of determination of the elementary composition of solid samples in range of 0.1-100% for elements in range of atomic number 12<Z<92. The XRF gamma spectrometers are mounted on vertically moving console of a commercial 3D printer. The sample is fixed on an x-y stage moving into horizontal directions. There is significant difference in the sensitive energy ranges of these two types of energy-dispersive semiconductor detectors (CZT and SDD); therefore it is possible to apply them simultaneously measuring the quantity of the major elements in the matrix and the activity of different isotopes. The XRF device has been constructed in confocal measuring arrangement, with approximately 2 mm of the focal spot diameter. Analytical capability, i.e. limit of detection, and lateral resolution of the confocal XRF spectrometer were determined. In order to evaluate and quantify the gamma spectra a new reverse Monte Carlo conception was designed and tested with application of the MCNP-6 validated code for simulation of gamma spectra recorded by a CZT detector having 1500 mm3 active volume.AcknowledgmentsThis work has been carried out in the frame of VKSZ-14-1-2015-0021 Hungarian project supported by the National Research, Development and Innovation Found.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 55 10 – 14 October 2016, Helsinki, Finland

SOURCE TERM DETERMINATION FOR VVER-440/V230 REACTOR DECOMMISSIONING BY MCNP5Martin Oravkin, Kristína Krištofová, Gabriel Farkas, Vladimír SlugeňSlovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovicova 3, 812 19 Bratislava, Slovakiamartin.oravkin@stuba.skABSTRACTThe main objective of the paper is to present calculated neutron source term results for a selected VVER440/V230 core configuration representing the first 12 campaigns of the V1 NPP at Jaslovské Bohunice site. Non-profiled fuel assemblies with steel spacer grids and relevant spatial burnup and coolant temperature distributions were taken into account for the campaign as well as the other integral core parameters, based on operational records.Transport Monte Carlo code MCNP5 with ENDF/B-VII.1 cross section libraries was applied to calculate the neutron source term on the core boundary and its characteristics.The determined source term is planned to be used for future transport calculation in order to evaluate the reactor induced activity (internal and external structural components) for the purpose of its decommissioning. The final major task will be a validation of calculated results with the experimental ones. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 56 10 – 14 October 2016, Helsinki, Finland

COMPARISION OF DETERMINED MCNP NEUTRON SOURCE TERMS FOR VVER-440/V230 DECOMMISSIONINGKristína Krištofová, Gabriel Farkas, Martin Oravkin, Vladimír SlugeňSlovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovicova 3, 812 19 Bratislava, Slovakiakristina.kristofova@kind-consultancy.comABSTRACTThe paper presents a comparison of different neutron source terms, determined for given VVER440/V230 core configurations, related to operational history of the V1 NPP at Jaslovské Bohunice site. The main objective was to determine the influence of diverse core characteristics, such as loading pattern, fuel assembly (FA) types, and presence of shielding assemblies (SA) on given source term (spatial, energy, and angular distribution of neutron flux at surface source boundaries). Four typical operational periods were distinguished: non-profiled FAs with steel spacer grids, non-profiled FAs with loaded shielding assemblies, non-profiled FAs with zirconium spacer grids and SAs, and profiled FAs together with SAs. Corresponding spatial burnup and coolant temperature distributions were taken into account for the selected campaigns as well as other integral core parameters, based on operational records. Transport Monte Carlo system MCNP5 with ENDF/B-VII.1 cross section libraries was applied to calculate the neutron source term on the core boundary and its characteristics.This resulted to a selection of reference core characteristics and relevant number of neutron source terms to be used in the following transport calculation and subsequently to evaluate the reactor activation for the purpose of its decommissioning. The final goal is to validate calculated results with experimental samples and use the validated model for VVER 440/V213 activation calculations, as well.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 57 10 – 14 October 2016, Helsinki, Finland

NUCLEAR CRITICALITY SAFETY ANALYSIS OF WET INTERIM STORAGE POOL USING MCNP5Gabriel Farkas, Vladimír Slugeň, Branislav Vrban, Jakub LüleySlovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovicova 3, 812 19 Bratislava, Slovakiagabriel.farkas@stuba.skABSTRACTThe paper presents results of nuclear criticality safety analysis of wet interim storage pool, located at Jaslovské Bohunice site. One of four pools loaded with non-compact T-12 respectively compact KZ-48 casks was modeled in MCNP5 code. The storage casks were filled with the II. Gen 4.87 % enr. fuel assemblies at different burnup levels. Conservative approach was applied and calculation of keff values was performed for normal and various postulated emergency conditions in order to evaluate the final maximal keff values. The requirement of current safety regulations to ensure 5% subcriticality was met in all investigated conservative cases. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 58 10 – 14 October 2016, Helsinki, Finland

DEVELOPMENT IN STUDSVIK’S SYSTEM FOR SPENT FUEL ANALYSESTeo Simeonov, Charles Wemple Studsvik Scandpower Inc teodosi.simeonov@studsvik.com charles.wemple@studsvik.comABSTRACTThe ongoing development of the system for spent fuel analyses at Studsvik aims to meet and fulfill a broad spectrum of backend reactor analyses, ranging from the very basic – isotopic concentrations, residual heat and activities – to support for cask loading procedures and accident analyses. The success of the SNF code among the utilities, research institutions, and national authorities is due to its best estimate methodology, combined with incomparable user friendliness, elaborate engineering functionality, and the ability to compute the spent fuel characteristics in real time over an entire reactor core and reactor spent fuel pool. Studsvik’s approach to spent nuclear fuel analyses combines isotopic data calculated by the CASMO5 or HELIOS2 lattice transport codes, with core irradiation history data from the 3D reactor simulator, and isotopic decay data. The overview in this paper aims to demonstrate the enhanced applicability of SNF as result of extensions and improvements in the methodologies in all three directions: lattice physics transport calculations, irradiation history, and decay data.KEYWORDS: spent nuclear fuel, decay library, ENDF/B-VII, SNF, CASMO5, HELIOS2 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 59 10 – 14 October 2016, Helsinki, Finland

INVENTORY OF SPENT FUEL VVER-440 WITH COOLING TIME UP TO 80 YEARSVladimír CHRAPČIAK, Radoslav ZAJAC VUJE Inc., Okruzna 5, SK 918 64 Trnava, Slovakia Vladimir.Chrapciak@vuje.skABSTRACTFrom safety point of view during storage and transport of spent fuel are important features such as decay heat, activity, neutron and gamma sources. Those parameters depend on initial enrichment, burnup and cooling time. In this article is comparison of basic parameters for Slovakia’s fuel for cooling time up to 80 years. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 60 10 – 14 October 2016, Helsinki, Finland

STUDY ON THE IMPACT OF TRANSITION FROM 3-BATCH TO 4-BATCH LOADING AT LOVIISA NPP ON THE LONG-TERM DECAY HEAT AND ACTIVITY INVENTORYTuukka LahtinenFortum Power and Heat Ltd, Espoo, Finland tuukka.lahtinen@fortum.comABSTRACTThe fuel economy of Loviisa NPP was improved by implementing a transition from 3-batch to 4-batch loading scheme between 2009 and 2013. Equilibrium cycle length, as well as all process parameters, were retained unchanged while the increase of fuel enrichment enabled to reduce the annual reload batch size from 102 to 84 assemblies.The transition had an impact on decay heat and activity inventory. This effect has not earlier been studied properly, i.e. by applying consistent calculation models and detailed description of assembly-wise irradiation histories. In this paper, the effect is analyzed with the emphasis on decay heat, produced by annual discharge batch, on the time scale which is relevant in final disposal studies.The transition involved a few changes having opposite consequences and, therefore, the net effect on decay heat is not self-evident. These changes include: modification of fuel assembly design decrease of uranium mass of reload batch (-13 %) increase of average burnup of discharge batch (+15 %) changes in assembly-wise end-of-life fission ratesThe paper documents the decay heat calculation results for both loading schemes. Based on the results, it is concluded that for the cooling time, foreseen typical prior to encapsulation of assemblies, the decay heat of discharge batch increases by 2…4 %. The net effect is dominated by increase of Pu-238 and Cm-244 activities. Furthermore, considering assembly-wise average decay heat, it turns out that prolongation of cooling time by about 10…15 years is needed when unchanged decay heat load is desired.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 61 10 – 14 October 2016, Helsinki, Finland

DEVELOPMENT OF A SOURCE TERM RELATED TO THE GEOLOGICAL DISPOSAL OF SPENT NUCLEAR FUEL IN SLOVAKIADana Barátová, Branislav Vrban, Vladimír NečasInstitute of Nuclear and Physical Engineering, Faculty of Electrical Engineering and Information Technology, Slovak University of Technology in BratislavaEmail: dana_baratova@stuba.sk, branislav.vrban@stuba.sk, vladimir.necas@stuba.skABSTRACTIn Slovakia, the preferred option how to manage a spent nuclear fuel (SNF) is a direct disposal of the SNF in a geological repository. It is considered that in the geological repository there will be disposed spent fuel assemblies from the operation of the Slovak nuclear power plants and also the radioactive waste which is not suitable for a disposal in the National Radioactive Waste Repository in Mochovce. Because of the fact that the isotopic composition of spent nuclear fuel and corresponding radionuclide activities are important parameters when developing a source term model of the geological repository, the isotopic composition and activities of individual radionuclides were calculated. Within this assessment the initial inventories of SNF assemblies were calculated for the fuel from VVER-440 reactors with different average initial enrichments of U-235 and different burn-up levels. Safety relevant radionuclides were identified based on the certain assumptions as relatively long half-lives, not negligible inventories, a presence of the instant release fraction (fraction of the inventory which is after water-contact released rapidly) and by using the international safety cases. Daughter nuclides with relatively short half-lives were assumed to be in equilibrium with their parent nuclides and were excluded from the radionuclide migration calculations. Stable isotopes were also included in the calculations as the isotopes of the same element share the solubility limit (the effect of reducing the solubility of safety relevant nuclides). Consequently the impact of different inventories of SNF assemblies on release rates (activity rates) in the radionuclide migration calculations was examined. Depletion calculations were realised by using the TRITON sequence of SCALE system and GoldSim simulation software was used for the radionuclide transport calculations. 26th Symposium of AER on VVER Reactor Physics and Reactor Safety 62 10 – 14 October 2016, Helsinki, Finland

EXTENSION OF POOL AT THE REACTOR ANALYSES IN SLOVAKIA AFTER FUKUSHIMARadoslav ZAJAC, Vladimír CHRAPČIAK VUJE Inc., Okruzna 5, SK 918 64 Trnava, Slovakia Radoslav.Zajac@vuje.skABSTRACTSlovak nuclear industry consists of four VVER-440 units in operation, interim spent fuel storage facility, two VVER-440 units and one A1 unit under decommissioning.Neutron Physical characteristics for routine operation are evaluated for each unit in routine operation. One of the neutron physical characteristics component is the chapter focused on support solutions of the emergencies states for safety state recovery or functional recovery of selected parameters.Chapter of design extension conditions analyses has been gradually integrated into the document of neutron physical characteristics after Fukushima accident according to Slovak NPPs and Slovak Nuclear Regulatory Authority requirements.Operator of the reactor unit is usually provided with data in tabular form, graphical dependences of core parameters and also for downtime period of reactor.The operator can read parameters when necessary:1. Decay heat of the core after reactor shutdown from the full power in the case of BOC and EOC. The history of decay heat is divided into four time ranges. The first one is set from 0 s to 120 s, the second range takes from 1 min to 60 min, the third from 1 hour to 24 hours and the last one from 1 day to 30 days.2. Decay heat of the core during the downtime period between previous and following cycle. The history of decay heat is divided into four time ranges. The first one is set from 0 s to 120 s, the second range takes from 1 min to 60 min, the third from 1 hour to 24 hours and the last interval depends on the length of reactor downtime.3. Minimum water flow through the core necessary to remove decay heat during the full power period and also during downtime period. The history of minimum flow is again divided into four time ranges. The first one is set from 0 s to 120 s, the second range takes from 1 min to 60 min, the third from 1 hour to 24 hours and the last interval is different and depends on the case of full power period or downtime period.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 63 10 – 14 October 2016, Helsinki, Finland 4. Decay heat in pool at the reactor dependence on the time. The graph represents actual sum of decay heat assemblies in the pool during the reactor period. The number of spent fuel assemblies varies from period to period. The time axis (x axis) is divided into months and the specific dates are written in axis.5. Time to achieve the limit parameters. This part follows three parameters dependant on time during the period. The first parameters present the time values to achieve limit coolant temperature 100 °C; the second parameter is the time to fuel uncovering and the third shows the time values to reach limit fuel temperature 1200 °C. The time axis (x axis) is divided into months and the specific dates are written in axis.6. The next part is focused on dependences of time to achieve the temperature of water saturation 100 °C in the pool at the reactor at the BOC and EOC, time of the lowering water to the level of fuel in the pool at the BOC and EOC, time to achieve fuel limit temperature 1200 °C at the BOC and EOC in the pool.7. The last point of neutron physical characteristics document is formed by two tables which define the point of boiling time in the reactor and in the pool.Before the Fukushima accident the document of neutron physical characteristics included time dependence of decay heat in the reactor at the BOC and EOC only. In the area of design extension conditions was defined a large series of analyses which became a serious part of neutron physical characteristics for VVER-440 reactors in Slovakia.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 64 10 – 14 October 2016, Helsinki, Finland