Band 85, Heft 4

The BIPR-8A executable module is a main component of the BIPR 2007 code. This module is intended for three-dimensional coarse mesh calculation of the VVER reactor cores. This paper presents the results of validation of the BIPR-8A executable module. Validation is based on comparison of the calculated results with the operating data of power plants with the VVER-440, VVER-1000 and VVER-1200 reactors, and with the results of calculations obtained by using Monte Carlo codes. The results of comparison of the following values at HZP (hot zero power) power level and operational power level are presented: critical concentration of boric acid, temperature reactivity coefficient, efficiency of the control rods system, relative power distribution of fuel assemblies and 3D power distribution, effective multiplication factor.

Studsvik’s in-core fuel management code package CMS5- VVER, which includes the CASMO5-VVER lattice physics code and SIMULATE5-VVER three-dimensional nodal code, is currently in use for VVER-1000/1200 reactor analysis. Recently, a new commercially available CASMO5 nuclear data library has been generated based on the ENDF/B-VIII.0 evaluation. The ENDF/B-VIII.0 evaluation represents the state-ofthe- art in nuclear data and features new incident-neutron cross section evaluations from the CIELO project for 1 H, 16 O, 56 Fe, 235 U, 238 U and 239 Pu. A summary of the main features and validation of the new ENDF/B-VIII.0-based data library, referred to as E8R0 library, is presented in this work. Comparisons of predicted criticality and fission rate distributions to measurements from various hexagonal-lattice critical experiments, such as the ZR-6 (TIC) and P-Facility, show excellent agreement between the E8R0-based calculations and measurements. In addition, validation results are presented for CMS5-VVER using the new E8R0 library and the X2 VVER-1000 benchmark problem. These results indicate that the E8R0 library provides comparable accuracy to E7R1 results for the various reactor physics parameters such as critical boron concentration, temperature reactivity coefficients, and control rod worth.

As long as VVER-440 reactors exist, their fuel design is subject to ongoing improvement. Efforts to find the optimal fuel design continue and encompass increasingly extending ranges of both geometry and material parameters of fuel assemblies. This paper presents the results of optimized radial enrichment profiling for RK3+ fuel assemblies and discusses whether it is absolutely necessary to use burnable absorbers (Gd 2 O 3 rods) in fuel assemblies suggested for prospective fuel cycles. Regarding RK3+ and RK2 fuel assemblies, this paper also presents thermohydraulic data yielded by computations performed using SC-1 certified software for a VVER-440 seven-assembly core fragment. These computations show that - compared to RK2 assemblies of the same capacity - RK3+ assemblies with optimized enrichment profile have more uniform radial temperature distributions and lower maximum outlet temperatures.

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

This work deals with the \Full-Core" VVER-1000 calculation benchmark which was proposed on the 26 th Symposium of AER [1]. Recently, the calculation benchmarks \Full-Core" VVER-440 [2] and its extension [3] have been introduced in the AER community with positive response [4, 5]. Therefore we have decided to prepare a similar benchmark for VVER- 1000. This benchmark is also a 2D calculation benchmark based on the VVER-1000 reactor core cold state geometry, explicitly taking into account the geometry of the radial reflector. The loading pattern for this core is very similar to the fresh fuel loading of cycle 9 at Unit 1 of the Temelin NPP (Czech Republic). This core is filled with six types of fuel assemblies with enrichment from 1.3%w 235 U to 4.0%w 235 U with up to 9 fuel pins with Gd burnable absorber per FA. 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. In this contribution we present the overview of available macro-codes results.

The VVER-1000 Full Core Calculation Benchmark was presented in the 26 th AER Symposium in 2016. Recently there are several solutions of this benchmark made by several reactor physics codes. In this paper the solution of the benchmark calculated by the KARATE code system will be presented in details and some comparisons to the reference solution will be shown.

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

Natural circulation (NC) of the coolant is an important component of ensuring safety and core cooling in conditions when forced flowrate has been lost. Unit 1 of Novovoronezh NPP-2 is the leading unit of the new NPP-2006 design with VVER- 1200. When commissioning this power unit, tests were conducted for the first time to verify the natural circulation mode, confirming both the development of this mode while simultaneously shutting down all the main coolant pumps (MCP), and as the design value of the permissible reactor power level in this mode. Due to the significance of the obtained experimental data during the commissioning tests and the need for validation of the calculation model for NPP-2006 design, numerical and experimental studies of the natural circulation mode were carried out jointly by the filial \Novovoronezhatomtechenergo" of the JSC \Atomtechenergo" and the National Research Center \Kurchatov Institute", taking into account their specializations. The analysis of experimental data demonstrates the ability of the plant to work in this mode. Calculation results obtained by ATHLET/BIPRVVER software package showed good agreement with the measured data, but a further improvement of the calculations of the NC mode is needed.

The paper describes the multi-physics coupling scheme between the SKETCH-N nodal neutronics code and the best-estimate thermohydraulic code ATHLET v3.2. Some first results are discussed. Various possible options of coupling have been considered. A scheme is selected and applied for data exchange between the codes based on MPI library. The verification and validation were performed using the transient of the Kalinin-3 international Benchmark. The simulation results show good agreement with experimental data and calculations performed by the participants of the benchmark. Parallel to the coupling scheme development, a visualization system to process the results is being created. The steady-state analysis is carried out using both simple and complex thermohydraulic models. The calculations of the transient \Switch off of one MCP (Main Coolant Pump) at nominal power" is performed applying a more elaborate thermohydraulic model taking into account inter- channel mass transfer.

This paper presents a test safety assessment calculation technique for operating high power VVER units in daily maneuvering regimes. The test of maneuvering operation is an actual problem of VVER reactors exploitation. Corresponding tests are classified as nuclear hazardous procedures which are accompanied by power variations, regulators actions and continuous space power field fluctuations. As a result a lot of possible reactor plant conditions can occur. This fact is taken into account during the safety assessment of such tests. The proposed technique is based on a conservative approach. The paper contains a step-by-step structure of the method, the description of the calculation procedures and the applied models for each step. Initial conditions for selected RIA accidents are given and justified. Conclusions important for the practice are supported by method application results analysis for assigned task. All calculations are conducted using the KORSAR/GP programming code with 3D-neutrion kinetic model.

Reactivity coefficients in respect to the fuel temperature and coolant density are the important integral parameters, which determine behavior of nuclear reactors in case of accidents. Experimental justification of these coefficients in the operating reactor is a quite complicated task. Changes of the coolant density and fuel temperature are the interdependent processes, which are difficult to separate. Experiments with excitation of axial and diametrical xenon oscillations, which were initiated by moving of the CPS CR (control rods of the control and protection system) group or an individual CPS CR, were carried in the framework the commissioning tests in the VVER-1200 reactor. Using the methods of mathematical statistics, we carried out multiple modelling of experiments with excitation of axial and diametrical xenon oscillations. Actual values of the reactivity coefficients, as well as some other parameters ensuring the best agreement of the calculated and experimental data were evaluated with the help of Bayes theorem.

A coupled thermal-hydraulics and reactor physics code system is being developed at the Institute of Nuclear Techniques of the Budapest University of Technology and Economics based on a higher-order transport approximation, the simplified spherical harmonics theory. The advantage of this method is that with a small increase in computational demand, it provides additional accuracy compared to diffusion theory. Besides - due to the fact that the multi-group SP 3 and diffusion equations have a mathematically similar form - it requires minimal effort to implement an SP 3 solution algorithm to an existing diffusion code. This paper focuses on an algorithm developed by the authors which applies Galerkin weighted residual method for spatial and theta method for time discretization. Results of two-group kinetic SP 3 calculations performed with the SPNDYN code are also presented for various one-dimensional perturbations taking into account the delayed neutron precursor balance equations as well. The flexible nature of the SP 3 equations makes the developed code a good starting point for more realistic dynamic calculations in the future.

Thermocouple located near the coolant outlet of the fuel assembly head may indicate a temperature biased by radiation heating and subsequently affect accuracy of incore measurements. The analysis splits into two parts; coupled neutronphoton Monte Carlo transport to determine radiation heating that serves as the input to heat transfer calculation resulting in temperature bias evaluation. Various nuclear reactions contribute to radiation heating in the active volume of the thermocouple steel. Monte Carlo transport calculation of radiation heating for average fuel assembly in one of latest Dukovany NPP cycles was performed with MCNP code. Heating power was compared to values published previously for Paks and Kola NPPs. Heat transfer calculations were performed with TEPLO code developed in SǨODA JS. Radiation heating can be considered as negligible if the axial shift between thermocouple active part and thermocouple socket is below 15 mm.

An important task of the current stage of development of the existing nuclear power system is to find a configuration in which it comes to full self-sufficiency in fuel. A significant share of Russian nuclear power plants at the current stage and in the near future are thermal VVER reactors operating in the open fuel cycle. The first stage of the transition to fuel self-sufficiency can be considered the option of loading these reactors with REMIX and MOX fuel, which will improve the fuel supply indicators. The paper presents a comparison of efficiency indicators presented by the fuel component of the cost and material balance for the scenarios of nuclear power plant development with VVER reactors loaded with classical uranium and REMIX/MOX fuel. The costs of fuel handling stages were considered as variable initial parameters. The calculations were carried out using the STEM-NES software developed at NRC \Kurchatov institute".

The paper describes an approach to solve the problem of closing fuel cycle of light water reactors (LWR) trough the multiple uranium recycling. Since the second uranium recycle there is a problem of full return of the recycled uranium to LWR fuel cycle due to significant increasing of 232 U concentration and strong regulatory restrictions on its content. The paper shows that the use of a double cascade is able to provide almost complete return of recycled uranium to the fuel cycle with minor losses. The reason for the losses is the necessity to clean the recycled uranium from the 232 U isotope. We developed a multicriteria approach to choosing the optimal cascade scheme parameters. The efficiency of the algorithm is demonstrated by the example of enrichment of the recycled uranium of the second recycle. It is shown that not correctly selected optimization criteria of the double cascade enrichment scheme can lead to the fact that the loss of the 235 U isotope, provided with waste treatment from 232 U, will exceed its savings.