Volume 83, Issue 4

Studsvik's 3D, steady-state, nodal code SIMULATE5, which is widely used for square lattice geometries, has been recently extended to hexagonal geometry for the analysis of VVER 1000 and VVER 1200s. The lattice physics code CASMO5 is used for generating nodal cross sections data from single assembly calculations. SIMULATE5's HEX extension solves the multi-group diffusion equation, or optionally SP3 equations, for the hexagonal-z geometry by dividing each hexagonal into six triangles. In each triangle, the Fourier solution of the wave equation is approximated by eight plane waves to describe the intra-nodal flux accurately. The cross sections are described by a hybrid microscopic-macroscopic model that includes approximately 60 heavy nuclides and fission products. Heterogeneities, due to spacer/grid and control rod, in the axial direction of an assembly are treated systematically. This paper summarizes the physics models employed in SIMULATE5-HEX as well as the initial benchmarking activities.

Discontinuity factors and node-wise group constants were generated by the SERPENT Monte Carlo code and applied in the KIKO3 DMG nodal code. The methodology was tested by calculating a typical VVER-440 calculation benchmark. A reference solution of the benchmark was calculated by using also the SERPENT code and the accuracy of the different approaches was checked against this latter solution.

This work deals with “Full-Core” VVER-440 extended calculation benchmark which was proposed on the 24 th Symposium of AER in October 2014 [2]. This benchmark is based on calculation benchmark defined by ŠKODA JS a.s. on the 21 st Symposium of AER in 2011 [1]. This benchmark differs from the first “Full-Core” VVER-440 benchmark in use of control rods from group No. 6. Reason why these benchmarks exist is problematic validation of power distribution predicted by macro-code on the pin by pin level against experimental data. This new benchmark is also a 2D calculation benchmark based on the VVER-440 reactor core cold state geometry with taking into account the geometry of explicit radial reflector. Loading pattern for this core is very similar to the first pattern of the Mochovce NPP. This core is filled with fuel assemblies with enrichment of 1.6%w 235 U, 2.4%w 235 U and 4.25%w 235 U. The main task of this benchmark is to test the pin by pin power distribution in fuel assemblies predicted by macro-codes that are used for neutron-physics calculations especially for VVER reactors. The reference solution has been calculated by MCNP6 code using Monte Carlo method and the results have been published in the AER community. The results of reference calculation were presented on the 27 th Symposium of AER in 2017 [3]. In this paper is presented comparison of available macro-codes results for this calculation benchmark.

The 2-D benchmark related to the “full-core” of VVER-1000 was proposed at the AER working group A&B meeting in 2015. Such type of benchmarks requires the adequate specification of the volume surrounding the core where geometry is complex enough. The proposed benchmarks have been solved using the 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.

The purpose of the work is to calculate VVER-1200 neutron-physical characteristics that are important for safety using MCU Monte Carlo code for an independent verification of the accuracy of design calculations of the first fuel cycle. The calculations were carried out before the startup of the first VVER-1200 power unit. A full-scale computer model of VVER-1200 was developed using the design documentation of the fuel assemblies and the reactor facility. Special attention was given to the accurate specification of the geometry and material description of the core and its immediate environment. The full-scale VVER-1200 model allows performing Monte Carlo calculation of some safety-relevant characteristics for which it is impossible or extremely difficult to carry out full-scale reactor experiments. In total, more than 110 different states were calculated; five states were calculated taking feedbacks into account. For all calculations, the neutron-physical characteristics obtained by means of the MCU code were used to verify the BIPR-7A and the PERMAK-A codes.

Throughout the existence of VVER-440 reactors, their fuel design was always subject to ongoing improvements. By the moment, extensive computations and RD&D have been performed to evaluate the feasibility of further improvement of VVER-440 fuel cycles and fuel design. This paper discusses computational assessments of basic neutronic parameters of prospective fuel cycles based on newly-designed fuel assemblies, namely RK3+, slim rod and RK3+SR. RK3+ stands for shroud less fuel assemblies with wide angle bars instead of shrouds and optimized pitches between fuel rods. Slim Rod means shrouded fuel assemblies with rods whose external diameter was reduced to 8.9 mm. RK3+SR is an abbreviation for shroud less fuel assemblies with wide angle bars instead of shrouds and rods whose external diameter was reduced to 8.9 mm. This paper demonstrates that, in the countries operating VVERs-440, a transition to new fuel design would improve both the energy output and the fuel cycle economy.

Recently, an initiative has been made to improve the fuel economy of VVER-440 reactors by implementing a modification on the geometry of the current fuel assembly design. The proposed modification involves reduction of the fuel rod outer diameter from 0.91 cm to 0.89 cm by using 0.01 cm thinner cladding tubes than earlier. The design improvement would shift the neutronics of an under-moderated system slightly towards optimum moderation and, therefore, increase the reactivity of the assembly. In this paper, a neutronics feasibility study on utilization of the proposed new fuel design at Loviisa NPP is carried out. The study involves a comprehensive comparison of two individual equilibrium fuel cycles: one applying current TVEL 2 nd generation fuel design and another one where the new fuel design is used. In addition to equilibrium cycle characteristics, also cycle economics as well as back-end effects are considered. The study concludes that the proposed fuel design modification enables to improve the fuel economy of Loviisa NPP.

Gen-IV fast reactors are envisaged to operate in closed fuel cycles due to their ability to breed their fuel from fertile feed and burn minor actinides produced by themselves or thermal reactors in the nuclear park. The optimization of such fuel cycle strategies requires detailed models, capable of simulating the transition from initial state to equilibrium. A fast and flexible burn-up scheme based on polynomial fitting of the one-group cross-sections, FITXS was used to develop burn-up models for the Gen-IV Gas-cooled Fast Reactor (GFR), Lead-cooled Fast Reactor (LFR) and Sodium-cooled Fast Reactor (SFR), as well as a European Pressurized Reactor (EPR) and a VVER-1200 MOX fuel assembly. The burn-up models were integrated in a closed fuel cycle model containing Gen-IV LFR and MOX fueled VVER-1200 reactors, and different scenarios were investigated and compared concerning the reduction of transuranium inventories and the stabilization of the plutonium inventory. Results show that the LFR is capable of burning minor actinides from spent VVER-440 fuel and that transuranium inventories can be stabilized or reduced with a mixed fleet of LFR and MOX fueled VVER-1200 reactors.

This article presents a comparison of results of simulation of VVER-440 reactor fuel burnup by means of ISTAR set of tools using ENDF/B-VII.0 and JEFF-3.1 evaluated nuclear data libraries. Code verification is required on the basis of both experimental data and computational uncertainty analysis. A VVER-440 fuel rod burnup simulation using ENDF/B-VII.0 and JEFF-3.1 evaluated nuclear data libraries was performed as preliminary step before comparison of nuclear data uncertainty analysis and calculation/experiment. Fuel irradiation conditions and SNF nuclide composition experimental data were obtained from radiochemical assays of irradiated VVER-440 fuel report. Fuel burnup simulation was carried out by means of ISTAR set of tools using Monte-Carlo code for spectrum averaged reaction rates evaluation required for solution of Bateman's equations. As a result, K inf and nuclide composition across burnup were estimated for the sample considered. On the basis of two nuclear data libraries’ implementation for depletion simulation it was shown that K inf uncertainties rise across burnup from 0.4% at the beginning of life up to 0.7% at the end of life. Heavy nuclides composition uncertainties run up to 9.5% for Am-241 at the end of the 4 th life time; however for main uranium and plutonium isotopes the uncertainties do not exceed 1.2%. Most fission products composition uncertainties stay below 7.3% except Sm-150 nuclide (15.1%). The next step will be a comparison with experimental spent nuclear fuel isotopic composition data considering the above mentioned uncertainties.

In 2016 OECD put forward a routine problem to simulate the operation conditions in VVER-1000 reactor of Rostov NPP unit No. 2 taking into account control rod motions. The participants were encouraged to simulate the integrated process and obtain the indications of the hot leg temperature control sensors depending on control rod motions. The data from in the experiment at 75% nominal power was selected for the first test carried out in the frames of the international task. Insertion of the 10th and 9th CPS rod groups into the core without driving was carried out in the low power testing experiments at Rostov NPP unit No. 2 (May 31, 2010). The task was split into several subtasks, each with its own specifics. One of such subtasks is focused on the simulation of the standard temperature control indications at the FA outlet. The way of solving the abovementioned subtask is reflected in the present report.

In Paks Nuclear Power Plant the VERONA core monitoring system is being used. In this article there is a new way presented to determine the precision of VERONA calculations during power transients. In that case VERONA uses a sophisticated method to follow the changes of the core. During a transient many online measured parameters cannot be used to determine the precision of the model due to their insufficient accuracy. However the change of the calculated boric acid concentration can be used for this purpose. If there is no inlet of boric acid or clean condense during the transient, at the end, the calculated boric acid concentration value should be equal to that, which was determined at the beginning of the power transient. However there is often a difference between these two values, which difference denotes the precision of the calculating model. This precision is heavily influenced by the heat conduction model, used between the fuel and the coolant. During the work five different conduction models were investigated, to determine which one offers more precise results. There was also a way examined to determine the uncertainty of the fuel temperature coefficient based on the outcome of the examined power transitions.

Physical startup tests are an important part of the operation of each nuclear reactor fuel cycle. This paper is focused on calculations of these tests for Dukovany NPP using MOBY-DICK diffusion macrocode. In this paper two tests are presented, specifically SCRAM test and “Weight of the 6 th group of control assemblies” test. Both tests are described, methodology of calculation is introduced and results of calculations are compared with experimental data.

The Refueling Neutron Monitoring System and the Reactivity Monitoring System for startup measurements are both aged, the development of a new combined system fulfilling both functionalities has therefore been started. The new system is based on 6 autonomous measurement chains covering the whole neutron flux range and uses fixed fission chambers. The signals of the measurement chains are received by two redundant processing systems continuously providing measured and calculated data for external systems like the VERONA core monitoring system. The system generates emergency signals and events for the operators of the Main Control Room and the Refueling Machine. The pilot measurement chain has been tested at Units 2 and 3 of Paks NPP. The first implementation of the new system is planned to be installed at Unit 3 by the middle of the following year.

In VVER-1000/1200 type reactors much higher and geometrically more complex assemblies are applied compared to the VVER-440 design. Additionally, assembly shrouds are not used in the improved reactor types. This implies that the existing sub-channel wise models and hot channel calculation methodologies used in the safety assessment of VVER-440 reactors have to be improved. For this purpose a full core thermal hydraulic model was developed. In the first part of the paper the new model and its connections with the reactor physics of KARATE-1200 code system are demonstrated with results regarding the sub-channel outlet temperature and DNBR. In the second part of the paper the hot channel calculation methodologies to be used in the transient safety analyses of VVER-1200 type reactors are discussed. The appropriate frame parameter concept concerning the limitation of the heating up of the coolant and the single closed sub-channel approach vs. the multi-channel approach are studied, as well.

This paper contains a description and evaluation of the thermal-hydraulic calculation of VVER-1000 transient connected with steam dump to atmosphere (SDA) opening during decreased reactor power to 20% of nominal power (N nom ). The calculation was performed with the thermal-hydraulic system program ATHLET coupled with the 3-D reactor dynamic code DYN3D. A comparison with the experiment was performed on the base of measured values during SDA project function test on the VVER-1000 Temelín NPP Unit 2. Results obtained from calculated vs. experimental values contribute to the validation of the ATHLET/DYN3D coupling.

A hypothetical main steam line break transient in a VVER-1000 core with multiple equipment faults was simulated with the coupled COBAYA4/CTF and COBAYA3/FLICA4 nodal core models. The objective was to test recent versions of the models developed in the EU NURISP and NURESAFE projects and to analyze the thermal-hydraulic conditions in the hot assembly at the pin-cell level. The accident scenario was specified as an aggravated variant of the OECD/NEA VVER-1000 MSLB benchmark with eight peripheral control rod clusters stuck out of the core after scram, all of them in the overcooled sector. Coarse-mesh models and a realistic cross-section library were used to compute the full-core behavior with pre-calculated MSLB boundary conditions and to identify the hot assembly. Then a sub-channel thermal-hydraulic analysis of the hot assembly was performed with the CTF code, using time-dependent assembly boundary conditions from the coarse-mesh vessel and core simulation. The results show that the predicted values of the safety parameters do not exceed the safety limits.