16th Symposium of AER on VVER Reactor Physics and Reactor Safety

Objectives for the “Program of modernization and safety upgrading of Bohunice V2” (1997) – safety, reliability, economy. Scope of technical activities in the project V2 .Modernization.

Mochovce NPP unit 3 and 4 is one of representants of delayed project in the world within nuclear energy area.
Mochovce NPP was originally planned for operation with four VVER-213 440 MW reactor blocks. Mochovce units 1 and 2 were commissioned in 1998 and 2000, respectively. Construction of Mochovce units 3 and 4 was suspended in 1992 due to a lack of financial resources. At present, construction of units 3 and 4 is approximately on 30% completed in terms of technology and approximately on 70% completed.
In September 1999, the Government of the Slovak Republic adopted Resolution No.801/99 to shut down the two reactor units of V-1 NPP at Jaslovske Bohunice in the years 2006 and 2008 pursuant to Slovakia’s accession negotiations with the European Union. As a result, 880 MW (out of 2,640MW) of nuclear capacity will be removed from SE’s assets basis. In addition, the expected closure of part of the production capacity from Novaky and Vojany thermal power plants (2006) and, later (around 2025), the scheduled closure of the two reactor units of V-2 at Jaslovske Bohunice NPP will lead to a large fall in SE’s (and Slovakia’s) installed (base load) capacity.
This has led SE to review the opportunity for completing the construction of reactor units 3 and 4 at Mochovce NPP. Under the proposed model of completion of Mochovce units 3 and 4, several fundamental changes in construction and technology from the operating Mochovce units 1 and 2 are planned. These include increasing the gross (net) installed capacity of Mochovce units 3 and 4 and extending the operating life of Mochovce units 3 and 4.
The Mochovce units 1 and 2 were selected as a reference plant with the following main areas for innovations and strategy of technical innovations of the original design solution of Mochovce units 3 and 4 were specified:
• nuclear safety corresponding to the safety level of the reference plant Mochovce units 1 and 2 and in conformance with current international requirements on nuclear and operational safety as a minimum, including handling of catastrophic accident issues,
• totally new instrumentation and control system for Mochovce units 3 and 4,
• brand new electric equipment according to STN IEC (Slovak technical norm The International Electrotechnical Commission) under preparation,
• possibilities for increasing production of electric energy,
• extension of lifetime up to 40 years of operation as a minimum,
• other possibilities for making future operation more effective.
This presentation deals with main safety, technical and economical issues shall be fulfilled to meet recent safety national and international requirements, as follows:
• Basic safety requirements on Mochovce units 3&4 design revision,
• Basic technical requirements on Mochovce units 3&4 design revision,
• Main phases of MO34 safety issues solution,
• Proposal for design changes – proposal for safety concept of Mochovce Units 3&4 construction completion,
• Safety measures within design basis accidents,
• Safety measures within beyond design basis accidents,
• Components and systems innovation,
• Increasing of technical and economical parameters of MO34.

Joint AER WG A and WG B held their fifteenth meeting in Modra–Harmónia, Slovak Republic during the period 24-25 April 2006.
The objectives of the meeting, content of presentations and future activities are shortly described in this paper.

The paper briefly describes inclusion of historical dependences of cross sections into MOBY-DICK code.
Changes in program and its library are specified, especially from the point of view of programs user.
Preliminary testing on assembly level is described and also testing on core level for “transient” loading pattern calculations is demonstrated on examples of 18th and 19th cycles of the Dukovany NPP Unit III.
Some features are addressed in the end of the paper

This article includes a test or in other words data verification of neutron ENDF/B-VIIb2 sub library. The first part consists from the process of preparation ACE files by NJOY 99.90. The starting point of data verification describes needed patches in NJOY 99.90, which are necessary to do for correctly production of ACE files. After the obtaining ACE files follow the test of all ACE files through GODIVA – input file for MCNP. GODIVA is high enrichment sphere of U-235, where every material is added as impurity. The aim of GODIVA test is to obtain a certainty if produced ACE files are able to run through MCNP.
The second part of this article begins with choose of benchmarks from “International Handbook of Evaluated Criticality Safety Benchmark Experiments, 2005”. From this source of criticality experiments were separated some benchmarks for quality verification of ACE files by MCNP.

Preparation of diffusion data for BIPR7 and PERMAK programs by spectral code HELIOS is described. Calculation sequence and HELIOS calculation tuning are characterised. Characteristics of the calculations are presented. Short comparison with Kurchatov institute results is given.

The report deals with fast and thermal neutron flows distribution in structural elements of VVER-1000 fuel assembly and absorbing rods, determination of absorbing isotope burn-up and worth variation in VVER reactor control and protection system rods.
Simulation of absorber rod burn-up is provided using code package SAPPHIRE_95&RC_VVER allowing detailed description of the core segment spatial model.
Maximum burn-up of absorbing rods and respective worth variation of control and protection system rods is determined on the basis of a number of calculations considering known characteristics of fuel cycles.

Unit 1 of Tianwan NPP with VVER-1000 reactor reached the first criticality at December 20 of 2005. Series of start up experiments were carried out under scientific advisory of RRC “Kurchatov Institute” specialists. At the Hot Zero Power state the reactivity coefficients, control rod group and scram worth were measured and symmetry of the core loading and power reactivity effect due to reaching 1% of nominal were assessed.
Paper describes in brief special features of experiments and presented some results obtained at a measurements.

Code package SAPPHIRE 95&RCVVER developed in FSUE NITI named after A. P. Alexandrov is used for VVER alternative neutronics calculations. The package allows providing full spectrum of calculations, from provision of small-group constants to calculation of fuel cycles and rod-by-rod power distribution.
The report deals with brief description of the computer package procedure and verification results for the application sphere covering core design calculations. The results of fuel cycle simulation are presented in comparison with operational data of VVER-440 power units.

Results of daily power’s maneuvering regime simulation that were proposed for 2nd unit of Khmelnitsky NPP are presented. Calculations were performed for campaign’s moments 100 and 150 fpd. Modeling was performed by spatial computer code DYN3D, that uses nodal method for neutron flux calculation. Calculated assessment of marginal value’s compliance of peaking factors, linear pin power, control rods efficiency and minimal axial offset deviation were performed. Comparative analysis of experimental data, that were obtained during maneuvering regime’s testing on 2nd unit of Khmelnitsky NPP, and modeling data are presented.

RRC “Kurchatov Institute” has performed coolant mixing investigation in a head of a full-size simulator of VVER-440 fuel assembly. The experiments were focused on obtaining the data important for investigating the trends in temperature difference between the value registered by a ICIS thermocouple and the value of average temperature. The completed experiments ensure representability of configuration simulation by reproducing every construction peculiar feature of flow part of fuel assembly in the domain between the lower spacing grid and thermocouple location, and also by slightly modified fuel assembly regular elements (or analogues thereof).
For the purpose of effectiveness of coolant mixing assessment within the head cross section of FA simulator, we measured coolant temperature distribution both in the place where coolant flow leaves the rod bundle simulator (in 39 data points along the cross section) and in the cross section location of regular ICIS thermocouple simulator (30 data points).
The testing was conducted with pressure of 90 – 95 Bar, mass coolant flow rates up to· 2000 kg/(m2⋅s), temperature of coolant heating in “hot” parts of the bundle up to 35оС and differences between coolant temperature extremes measured in rod bundle simulator outlet up to 20оС. Temperature fields were registered in 63 conditions that differ in coolant flow and inlet coolant temperature, electrical heating rate of FA simulator, and radial coolant distribution. In certain registered conditions we simulated coolant leakage to the space between the fuel assemblies.
The received test data may be important both for investigation of dependencies between the coolant temperature in regular thermocouple location or average outlet temperature in assembly head, and for validation of CFD codes or subchannel codes.

In the RRC “Kurchatov Institute” during 3 last years it was realized work package directed to development of the method of power assemblies correction using thermocouples readings.
This entire works include the experimental and calculation ones. The main result of the experiment investigations is fact of incomplete coolant mixing in the head of VVER-440 assembly. Furthermore experimental data were used for validation of CFD code PHOENICS and modernized subchannel code SC-1 (SC-1 is analogue of COBRA-IV code).
Modernized SC-1 code is capable to calculate the coolant temperature distribution in not only the fuel part of assembly but the head of assembly too.
In the future the modernized SC-1 code was used for calculation analysis of the thermocouples data that were obtained from 3-d and 4-th units of the Kola NPP.
At the results of this analysis were developed the method of the moving of incomplete coolant mixing correction in the nozzle of VVER-440 assembly.
The developed corrective functions were used for compare of calculated and measured relative assemblies power using thermocouples readings.
The main result of such comparison is that the differences between calculated and measured relative assembly powers for representative sample are reduced as a whole. Total amount of the considered variants were more than 1200 (different units, campaigns, assemblies and time moments).

The purpose of this work consists in investigation of the gadolinium fuel pin (FP) influence on space power distribution, especially from viewpoint of the values and gradient occurrence inside of neighbouring FPs that could result in static loads with some consequences, e.g., FP bowing. Since detailed power distributions cannot be obtained in the NPPs, needed information is provided by means of experiments on research reactors. As for the power release measurement inside of FPs, some special (e.g. track) detectors placed between fuel pellets are usually used. Since such works are relatively complicated and time consuming, an evaluation method based on mathematical modelling and numerical approximation was proposed by means of that, and using measured (integral) power release in selected FPs, relevant information about power release inside of needed (investigated) FP, can be obtained. For this purpose, an experiment on light water, zero-power research reactor LR-0 was realized in a WWER-1000 type core with 7 fuel assemblies at zero boron concentration and containing gadolinium FPs.
Application of the above evaluation method is demonstrated on investigated FP neighbouring a FP with gadolinium by means of the
• Azimuthal power distribution inside of investigated FP on their fuel pellet surface in horizontal plane and
• Gradient of the power distribution inside of investigated FP in two opposite positions on pellets surface that are situated to- and outwards a FP with gadolinium.
Similar information can be relevant from the viewpoint of the FP failures occurrence investigation.

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

Optimization of loading patterns is a very important task from economical point of view in a nuclear power plant. The optimization algorithms used for this purpose can be categorized basically into two categories: deterministic ones and stochastic ones. In the Paks nuclear power plant a deterministic optimization procedure is used to optimize the loading pattern at BOC so that the core would have maximal reactivity reserve.
To the group of stochastic optimization procedures belong mostly simulated annealing (SA) procedures and genetic algorithms (GA). There are new procedures as well which try to combine the advantages of SAs and GAs. One of them is called population mutation annealing algorithm (PMA).
In the Paks NPP we would like to introduce fuel assemblies including burnable poison (Gd) in the near future. In order to be able to find the optimal loading pattern (or near-optimal loading patterns) in that case we have to optimize our core not only for objective functions defined at BOC, but at EOC as well. For this purpose I used stochastic algorithms (SA and PMA) to investigate loading pattern optimization results for different objective functions at BOC and EOC.
In this paper (and at the symposium) I would like to present these results.

The new benchmark based on Dukovany NPP Unit-3 history of Gd-2 fuel type utilisation is defined. The main goal of this benchmark is to compare results obtained by different codes used for neutron-physics calculation. Input data are described in this paper including initial state definition. Requested output data format for automatic processing is defined. This paper includes:
a) fuel description
b) definition of starting point and five fuel cycles with profiled fuel 3.82% only
c) definition of four fuel cycles with fuel Gd-2 (enr.4.25%)
d) recommendation for calculation
e) list of parameters for comparison
f) methodology of comparison
g) an example of results comparison.

The improvement of nuclear fuel cycle economy results in permanent introducing of new types of fuel. The goal is to maximize the burn up without any increase the leakage of fuel elements. The next generation of fuel with burnable absorber has been operated in Bohunice V-2 NPP since this year. This paper summarizes in brief the whole process of introduction of new fuel generation with burnable absorber Gd and first experience with it.

At the end of 2002 Units 1 and 2 of Kozloduy NPP were shutdown before their design life time which left a large number of assemblies yet with a significant energy resources. A decision to load these assemblies into the cores of Units 3 and 4 during the next 4 cycles has been taken.
In 2003, 43 assemblies from Unit 1 cycle 23rd and 55 assemblies from Unit 2 cycle 24th are loaded in the cores of Units 3 and 4 respectively.
In 2004, new 49 assemblies from Unit 1 cycle 23rd and new 55 assemblies from Unit 2 cycle 24th are loaded in the cores of Units 3 and 4 respectively.
In 2005, the next new 25 assemblies from Unit 1 cycle 23rd and 66 assemblies from Unit 2 cycle 24th are loaded in the cores of Units 3 and 4 respectively.
In 2006, the next new 54 assemblies from Unit 1 cycle 23rd and 52 assemblies from Unit 2 cycle 24th + 2 assemblies from Unit 3 cycle 19th are loaded in the cores of Units 3 and 4 respectively.
The SPPSHB computer code system is used for development and safety assessment of the fuel loading patterns of Units 3&4 at Kozloduy NPP with low burned assemblies from Units 1&2.

Construction of VVER-440 fuel assemblies was being continuously modernized throughout the whole period of these reactors’ operation. The interest in further increasing the fuel cycle efficiency has currently resulted in the technical proposal of a third-generation assembly, which incorporates the following improvements:
• replacement of working assembly shroud with stiffening angles;
• reduced fuel element diameter (higher water-uranium ratio);
• reduced fuel cladding thickness from the inside (a possibility to load more uranium);
• increased diameter of fuel pellet with exclusion of its central hole (increased mass of uranium in the fuel assembly);
• replacement of several fuel elements in the working assembly with displacers of various kinds;
• increased fuel enrichment – up to 4.95% of uranium-235.
The paper summarizes the results of calculations of the basic neutron physics parameters for various configurations of VVER-440 third-generation RK-3 fuel assemblies, the results of computer exploration of fuel loads with various RK-3 options and the results of computer analyses of fuel element’s behaviour in RK-3 fuel assemblies. The paper also provides a technical and economic assessment of fuel cycles with different RK-3 fuel assembly options for VVER-440.

Compensation problem of 236U initial content in recycled uranium fuel for VVERs is considered. The calculation method for fuel compensative additional 235U enrichment is proposed. It is proved that a compensative enrichment can be well approximated by a linear function of 236U content in fuel. Approximation coefficients linearly depend on the effective fuel enrichment. After the compensation performed by the proposed method, the multiplication properties of fuel assembly with both natural and recycled uranium fuel show little difference in the course of a fuel cycle. Therefore, for the purpose of recycled fuel safety it is possible to utilize the same safety margin as in case of the natural uranium fuel.

The old designed Russian fuel assemblies TVS-M have been under operation for many years at Kozloduy NPP WWER-1000 reactors. It was gained a lot of experience with TVS-M assemblies operation and a distortion tendency was found out. To avoid this problem a decision for a gradually substitution of TVS-M fuel assemblies with advanced ones (TVSA) has been taken.
Two different stationary fuel cycles with 42 and 48 fresh TVSA fuel assemblies are presented and discussed in this paper.
The KASKAD computer code system is used for development of the fuel loading patterns of Units 5&6 at Kozloduy NPP.

AER Working Group C activity in 2006

Standard instrumentation systems at NPP are used for data acquisition, processing, transmission, archiving, etc. The standard systems carry out mentioned tasks with predetermined sampling period that should be constant during the time. Unfortunately, the real operating sampling interval does not have enough accuracy. If such standard system is used as a supplier of neutron flux for reactivity computer, the unstable sampling interval can influence calculated reactivity accuracy.
The influence of sampling interval inaccuracy on calculated reactivity error is analysed in this paper. At first the influence is investigated in the case of reactivity step change, then in the case of reactivity course obtained during reactor criticality achievement. On the basis of the gained results, the software module KRITEX in SCORPIO-Bohunice monitoring system that is used for monitoring of reactor criticality achievement was upgraded. The analysis results and some KRITEX test results are presented in the paper.

This paper encompasses such section as objective, conception and engineering solution for construction of advanced in-core instrumentation system for high power reactor, including VVER-1000.
The ICIS main task is known to be an on-line monitoring of power distribution and functionals independently of design programs to avoid a common cause error.
This paper shows in what way the recovery of power distribution has been carried out using the signals from in-core neutron detectors or temperature sensors.
On the basis of both measured and processed data, the signals of preventive and emergency protection on local parameters (linear power of the maximum intensive fuel rods, DNBR, peaking factor) have been automatically generated.
The paper presents a detection technology and processing methods for signals from SPNDs and TCs, ICIS composition and structure, computer hardware, system and applied software. Structure, composition and the taken decisions allow combining class 1E and class B and C tasks in accordance with international standards of separation and safety category realization.
Nowadays, ICIS-M is a system that is capable to ensure: monitoring, safety, information display and diagnostics function, which allow securing actual increase of quality, reliability and safety in operation of nuclear fuel and power units. Meanwhile, it reduce negative influence of human factor on thermal technical reliability in the operational process.

The paper analyzes and discusses possibility of neutron flux inspection in the VVER core during fast dynamic processes applying existing in-core monitoring system.
The structure and functions of the system, basic principal of detector functioning and its temporal parameters are described briefly.
To assess the ability of such dynamic monitoring the event with control rod drop happened during operation of Kozloduy NPP Unit 5 is observed – at the level of power close to nominal one of the rod from control group shifted to the lowest position at ~2 seconds. In-core detectors readings at the process were registered and processed with mathematical methods that allow to single out only the prompt part of the signal. Results of the processing are presented.
Furthermore, the process observing have been modeled with 3D dynamic code NOSTRA. Results of modeling are presenting in a paper, and comparing with experimental ones. A good agreement achieved.
The analysis of measurements and its imitation give a hope that with an aggregate signal of detectors the measurement of control rod worth could be provided, and it allows to avoid of influence of spatial effects that are significant at standard technique with ex-core ion chambers.

An accurate reconstruction of the pin-wise power distribution is an important part of the limit checking of reactor operation. In SCORPIO-VVER core monitoring system, this is provided by the RECON module on the basis of pre-calculated form-functions. A new technique was implemented for improvement of the accuracy of pin-wise power reconstruction via a new methodology of form-functions calculations.

In the near future at the Paks NPP, a project for upgrading the VERONA core monitoring system will have been finished. The necessity of the modernization of VERONA system was indicated by the reactor power up rating, introduction of new fuel types and the lifetime extension of units. With parallel the VERONA modernization a hand able and easy to use tool, called VERONA-expert, has been developed for reactor physicists. The VERONA-expert can fulfill the same functions as the VERONA system without the standard VERONA computer architecture. What’s more, the VERONA-expert capable to make a lot of additional analysis. One can find the reason of the development of VERONA-expert system in the duality in the on-line in-core analysis. During in-core analysis it is very important, on one hand, to make reliable analysis of the reactor core in on-line way with a completed and licensed system. It can be done by VERONA system. On the other hand, it is also very important to make fast and reliable analysis of new and unexpected phenomena in the reactor core in on-line or off-line way. It can be done by VERONA-expert system.
The presentation gives brief summery of main features of the VERONA- expert system.

In this a new type of in-core sensor validation method is outlined, which is based on combination of correlation coefficients and mutual information indices. First experience with this approach is described and further improvements to enhance the outcome reliability are proposed namely through computational inteligence techniques.

The contribution of fuel assemblies to the response of ex-core detectors does not only depend on the power, but also on the position of the given assemblies in the reactor core. The weight of the inner fuel assemblies is several orders of magnitude lower than the outer ones. Therefore, the signal of the ex-core detectors for a given reactor power is strongly influenced by the spatial power distribution and indirectly by the parameters which determine the distribution, such as load pattern, time elapsed, etc.
The aim of the work is to describe the modelling and the methodology for calculation of the spatial weight function of ex-core detectors for the WWER-440 reactor type using Monte Carlo transport code system MCNP-4C2, taking into account different operational parameters such as power, burn-up, boric acid concentration, position of the control assembly group No 6, etc. The spatial weight function provides relationship between the spatial neutron flux density distribution or the fission density distribution in the reactor core and the ex-core detectors. The weight function is possible to define in various ways in dependence on solved problem. In this case the function gives the average number of reactions occurred in the detector for one neutron born with Watt fission spectra created in the twentieth of a fuel pin in different reactor core position.
From the definition of the weight function follows that the response of the ex-core detector is given by the weight function and the fission density distribution function product and its integration over the reactor core or its part.
The spatial weight function will be used for interpretation of reloads startup rod drop measurements, assessment of the ex-core detectors response at deep subcritical reactor stage and for other applications.

Report describes results of utilization in practice knowledge about dynamic method theory. The paper shows values of static and dynamic coefficients necessary for correction signal. The conclusion recommends possibility for future.

Standard core monitoring and information systems are designed with an emphasis on normal operation of nuclear power plant. Their purpose is to provide necessary support for reactor operators and other operating personnel during the fuel cycle. After each fuel reloading, and to the larger extent during the plant commissioning, a variety of start-up tests need to be carried out and evaluated. Sampling periods, accuracy and communication delays of standard systems are not always suitable for test performance and evaluation. For technical and safety reasons, the access to standard monitoring and information systems is very limited.
Non-Standard Measurement and Evaluation Systems are highly specialized devices designed with an emphasis on start-up tests performance and evaluation. They are capable of high frequency sampling, processing and communication of hundreds of technological signals with required accuracy and low communication delay. All technological signals needed for the test performance and evaluation are collected from various systems, concentrated in one system and fully accessible to a test leader, a reactor physicist or other users. In addition, Non-Standard Measurement and Evaluation Systems can perform other tasks like data storing and presentation, data distribution to other systems or external computers via network, reactivity calculations, etc. Structure of such systems can vary, but generally it is a mixture of two basic concepts: a mobile system and a stationary system. A basic description of hardware structure and software equipment of Non-Standard Measurement and Evaluation Systems is given in the paper.

AER Working Group D on VVER reactor safety analysis held its 15th meeting in Grand Hotel Duomo in Pisa, Italy during the period 26-27 April 2006. The meeting was hosted by the University of Pisa following the fourth workshop on the OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark (V1000-CT) held at the same location on 24-25 April. Altogether 15 participants attended the Working Group D meeting, 11 from AER member organizations and 4 guests from non-member organizations. The coordinator for the working group, Mr. P. Siltanen (FNS) served as chairman. In addition to general information exchange on recent activities in the participating organizations, the topics of the meeting included:
• Code development and benchmarking for reactor dynamics applications.
• Safety analysis methodology and results.
• Future activities.
A list of participants and a list of handouts distributed at the meeting are attached to the report.

An intensive research has been started in our institute focusing on the hydrodynamics of fuel rod bundles. Both the numerical and the experimental works have been started. One of the several home-developed computational codes is based on the spectral element method. The spectral element method is a relatively new and promising method to simulate incompressible fluid flows in complex geometries. It unites the geometric flexibility of finite element methods and the accuracy of spectral methods. The aim of the work is to accurately simulate and explore the secondary flow patterns of fluid flow in a subchannel of VVER-440 fuel rod bundles.
The concept of the spectral element methods is presented, and the algorithm used to solve the Navier-Stokes equations is detailed. The design and implementation issues of a parallel spectral element code that is able to simulate fluid flows in arbitrary geometries are also discussed. Some preliminary results of flow simulations for a subchannel of a fuel rod bundle are presented. These results are compared with the results obtained by a code based on the lattice Boltzmann method.

The coupled code system ATHLET-BIPR8KN is being developed to perform realistic simulation of three-dimensional neutron kinetics and thermal-hydraulic processes in VVER reactors. The detailed description of the flow mixing in the all reactor vessel is of great importance for the correct modeling and simulation of many NPP transients. As contribution to the validation of the system code ATHLET-BIPR8KN for asymmetric core inlet conditions, the CEA-NEA/OECD VVER-1000 Coolant Transient Benchmark (V1000CT) – vessel mixing problems are analyzed.
Results are presented for Exercise 1 of the Phase 2 of the Benchmark obtained with the system code ATHLET-BIPR8KN, supplemented by sensitivitiy studies done on the variation of the number of the thermo-hydraulic channels with cross flows (CF) in down comers and bottom plenums under the core with CF and with 163 (1:1 mapping schema) parallel thermal-hydraulic channels (PTHC) without of CF in the core and 48 PTHC without of CF for the core bypass. The transient is a simultaneous isolation of one steam generator from the steam line and the feed water supply at low power level. It causes a temperature rise in the affected loop and leads to a non-uniform temperature distribution at core inlet. The results of the simulation are compared with experimental data from NPP Kozloduy Unit 6 collected during the plant-commissioning phase.

In the German practice of considering boron dilution transients (BDT) in safety analysis reports (SAR), a strongly conservative approach is applied, although it is not explicitly requested in German rules and guidelines for SAR. The approach is based on recommendations of the German Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) and the technical expert organizations (TÜV) and accepted by the Reactor Safety Commission (RSK) and is currently followed by the industries and the facilities . No final recommendation and guidelines exist, because the item is subject of comprehensive discussions and research in Germany currently.
This conservative approach is based on a combination of analytical and experimental steps. First of all, in a series of thermal hydraulic system code calculations, the bounding scenarios are determined which lead to maximum realistic (under the conditions assumed for nuclear safety verification calculations) volumes of the lower-borated coolant which can be transported to the reactor core by re-establishing circulation in the primary circuit. These are mainly small break loss of coolant accidents (SBLOCA), during which the decay heat is removed from the core by the reflux-condenser regime. The position and maximum size of lower-borated slugs are identified, and the circulation re-start conditions are determined. The main SBLOCA scenarios are verified on experiments at large scale test facilities (e.g. PKL in Germany). While in the integral tests, the formation and transport of the slugs in the loops is assessed, the mixing is investigated in detail in dedicated mixing test facilities (ROCOM) using the boundary conditions either from the integral tests or from the system code calculations. In this way, the minimum boron concentration at the core inlet, which is reached during the transient, is estimated. Additionally, the minimum boron concentration is calculated by using computational fluid dynamics (CFD) codes. The critical boron concentration of a reactor core should be lower than the minimum boron concentration at the core inlet, determined in the mixing experiments or obtained from reliable calculations. This approach contains a high amount of conservatism, because it is assumed that the boron concentration is uniform in the whole reactor core. The spatial and temporal distribution of the boron concentration in the reactor core, available from the detailed analyses, is not taken into account.
A best-estimate approach has been developed at FZR and was applied to generic studies of BDT scenarios connected with start-up of the first MCP for German KONVOI type reactors. This approach starts with the definition of a scenario which covers all possible BDT scenarios with respect to reactivity consequences (maximum reactivity insertion). This is followed by a reactivity initiated accident (RIA) analysis using appropriate best-estimate tools (coupled 3D neutron kinetics/thermal hydraulic codes). Best-estimate boundary conditions for the boron concentration at the core inlet are applied. These are time-dependent boron concentration at the core inlet determined from experiments or CFD calculations. The final goal of the analysis is to show the integrity of the fuel rods. Usual acceptance limits with respect to maximum fuel temperature, maximum cladding temperature, radially averaged enthalpy deposited in the fuel and maximum cladding oxide layer thickness for RIA are applied.
The use of this best estimate approach, which should be accompanied by an uncertainty analysis, can help to make the design of future reactor cores more flexible and economically not decreasing the safety level of the nuclear installations.

The fulfillment of the safety analysis acceptance criteria is usually evaluated by separate hot channel calculations using the results of neutronic or/and thermo hydraulic system calculations. In case of an ATWS event (inadvertent withdrawal of control assembly), according to the analysis, a number of fuel rods are experiencing DNB for a longer time and must be regarded as failed. Their number must be determined for a further evaluation of the radiological consequences.
In the deterministic approach, the global power history must be multiplied by different hot channel factors (kx) taking into account the radial power peaking factors for each fuel pin. If DNB occurs it is necessary to perform a few number of hot channel calculations to determine the limiting kx leading just to DNB and fuel failure (the conservative DNBR limit is 1.33). Knowing the pin power distribution from the core design calculation, the number of failed fuel pins can be calculated. The above procedure can be performed by conservative assumptions (e.g. conservative input parameters in the hot channel calculations), as well.
In case of hot channel uncertainty analysis, the relevant input parameters (kx, mass flow, inlet temperature of the coolant, pin average burnup, initial gap size, selection of power history influencing the gap conductance value) of hot channel calculations and the DNBR limit are varied considering the respective uncertainties. An uncertainty analysis methodology was elaborated combining the response surface method with the one sided tolerance limit method of Wilks. The results of deterministic and uncertainty hot channel calculations are compared regarding to the number of failed fuel rods, max. temperature of the clad surface and max. temperature of the fuel.

The paper describes methodology for quantification of fuel cladding failure as a result of Loss of Coolant Accident (LOCA) for VVER-440 reactor type. The methodology is based on external coupling of the thermo-hydraulic code RELAP5 and thermo-mechanical code TRANSURANUS. The thermo-hydraulic response of the unit to the accident is simulated by RELAP5 code, providing initial and boundary conditions for the thermo-mechanical simulation by the TRANSURANUS code. Cladding failure criterion of the TRANSURANUS code, derived and implemented into the code in the framework of EXTRA EURATOM 5th Framework Programme is used. Cladding failure probability is evaluated by the Monte Carlo algorithm varying the outer cladding temperature. In the second part of the paper, an example of application of the methodology for typical maximum design accident of the VVER-440 is given, presenting every step of methodology and typical failure rate for this type of accident.

Results of thermo-mechanical behavior of fuel pins during daily power’s maneuvering regime that were proposed for 2nd unit of Khmelnitsky NPP are presented. Calculations were performed for campaign’s moments 100 and 150 fpd and for different type of regulation. Calculations of macro-core parameters (Kq, Kv) was performed by spatial computer code DYN3D. Calculations of micro-core parameters (fuel pin power) was performed by computer code DERAB. Calculations of thermo-mechanical behavior of fuel pins was performed by computer code TRANSURANUS.

Power upgrading of VVER-440 type reactors are being planned in several countries. From the point of view of both the planning and the licensing the detailed knowledge of the thermal-hydraulics processes in the fuel assemblies is essential. In the last years the evolution of the commercial CFD codes and the growth of the computer capacities have enabled the application of the CFD codes in this field.
Towards to get to know the thermal hydraulics processes in a VVER-440 fuel assembly the development of a complex fuel bundle model is under way in the Institute of Nuclear Techniques of the Budapest University of Technology and Economics. In the paper our VVER-440 fuel bundle models and results of our calculations were presented.
At first a 240 mm long and – considering the symmetries – 60 degree segment model of fuel bundle was developed using the ANSYS ICEM software. The model geometry contains the 240 mm length part of 24 fuel pins, the central tube, the inner wall of the shroud and one spacer grid. Four different fine meshes were developed and wide mesh sensitivity studies were performed.
Using the 240 mm long model a full length (2420 mm) fuel bundle model was built, which involves the 24 fuel pins, the central tube, the inner wall of the shroud and ten spacer grids. The model contains near to 7 millions volume elements. With this model calculations were carried out using ANSYS CFX 10.0 CFD code. In the calculations a non-profiled fuel assembly of 3.6% enrichment was investigated which was in operation in the Paks NPP.
The computations were accomplished using several RANS turbulence models and the effects of choosing different turbulence models were in details investigated.
For the same assembly calculation was performed with the COBRA subchannel code too. The results of the CFX and the COBRA calculations were also compared.

In the framework of the VVER-1000 Coolant Transient (V1000CT) Benchmarks coolant mixing and main steam line break (MSLB) benchmarks for VVER-1000 reactors were defined. Reference plant is Kozloduy-6, measured data from NPP mixing experiments are available to validate the calculation results, and – within the benchmark program – code-to-code comparison is to be used to evaluate the reactor vessel thermal hydraulic models. As a part of this program an exercise is specified to test the capability of vessel thermal-hydraulic models to represent the vessel mixing. Plant measured data from VVER-1000/V320 coolant mixing experiments are available to test and validate vessel mixing models (CFD, coarse-mesh and mixing matrix). The task of the benchmarks is to compare the calculations with measured data, using specified vessel boundary conditions and core power distribution. The measurements were conducted during the plant commissioning phase at Kozloduy-6 and Kalinin-1, 2. CAD geometry model of the reactor pressure vessel is provided as a part of the source information for the plant data.
In the present paper mixing calculations based on the V1000CT benchmark definition will be presented. For the calculations the ANSYS CFX-10 three-dimensional Computational Fluid Dynamics code was used. The calculations cover steady state analyses of the final state of the Kozloduy-6 mixing test. In this state on cold leg No. 1. the coolant temperature is higher (282.2 °C) than the temperature of other cold legs (cold leg No. 2.: 269.9 °C, No. 3.: 269 °C and No. 4.: 269.2 °C). Measured fuel assembly outlet temperatures, fuel assembly mixing coefficients and estimated fuel assembly inlet temperatures are provided. For the modeling of the VVER-1000 reactor pressure vessel non-structured tetrahedral mesh was used. The mesh was generated by the code ICEM using the provided CAD geometry of the reactor pressure vessel. The results of the CFX calculations will be compared with the available measured and estimated plant data.

The paper summarises present works performed by VUJE, Inc. in the field of CFD simulation of coolant flow in fuel assemblies used in the VVER-440 reactors. These works are extension of the previous calculations presented on AER Symposium 2004 and their aim is the better understanding of coolant flow in different parts of fuel assembly. While the previous CFD simulations were focused on study of individual parameters influences on the coolant flow mixing in the upper part of fuel assembly (pin-wise power distribution, geometry of fuel assembly, by-pass flow and coolant flow in central tube), the present works are directed more to verification of previous calculations and joining of individual effects on coolant flow mixing in the upper part of fuel assembly, mainly in the thermocouple position. The main purpose of the CFD simulations of coolant flow in fuel assembly is to increase accuracy of temperature measurements placed at the output of fuel assemblies.

AER Working Group E in 2006
Vladimir Chrapciak, VUJE, Inc., Slovakia, chrapciak@vuje.sk

In this article are analyzed experiments for VVER-440 fuel and compared with theoretical results by new version of the SCALE 5 code:
nuclide compositions – measurement in Kurchatov institute for 3.6%
- measurement in Dimitrovgrad for 3.6% (project ISTC 2670)
The focus is on modules TRITON and ORIGEN-S.

KARATE is a code system developed in KFKI AERI. It is routinely used for core calculation. Its depletion module are now tested against the radiochemical measurements of spent fuel samples from the Novovoronezh Unit IV, performed in RIAR, Dimitrovgrad. Due to the insufficient knowledge of operational history of the unit, the irradiation history of the samples was taken from formerly published Russian calculations. The calculation of isotopic composition was performed by the MULTICEL module of program system. The agreement between the calculated and measured values of the concentration of the most important actinides and fission products is investigated..

In the paper some results of validation of the SCALE4.4 depletion control module SAS2H against experimental data are presented. The recently published ISTC report with VVER-440 PIE data for fuel samples from Novovoronez NPP and ORNL publication with VVER-1000 experimental data for fuel samples from Balakovo NPP have been applied for preliminary validation of depletion calculations. The calculated results for 14 actinides and 14 fission products for VVER-440 samples and for 12 actinides and 5 fission products for VVER-1000 samples have been compared with measured isotopic concentrations.

Aiming at validation of depletion codes against WWER-440 spent fuel data some calculations of isotope composition of WWER-440 spent fuel assembly have been carried out by the NESSEL-NUKO code system. The initial data and data for the comparisons were taken from the ISTS burn up credit project data, recently published in the ISTC report ”Radiochemical Assays of Irradiated WWER-440 Fuel for Use in Spent Fuel Burnup Credit Activities.
The specific work scope included the destructive assay (DA) of spent fuel assembly rod segments with an ~38.5 MWd/KgU burn up from a single WWER-440 fuel assembly from the Novovorenezh reactor in Russia.

Improved calculational methods allow one to take credit for the reactivity reduction associated with fuel burnup. This means reducing the analysis conservativism while maintaining an adequate criticality safety margin. Application of burnup credit requires knowledge of the reactivity state of the irradiated fuel for which burnup credit is taken. The isotopic inventory and reactivity has to by calculated with validated codes.
By depletion calculation is necessary to take into account:
• List of used nuclides
• Cooling time
• Core parameters by irradiation
- Fuel temperature
- Moderator temperature/density
- Soluble boron
- Specific power and operating history
• Axial and horizontal burnup profiles
In this paper is numerical evaluation of impact of above parameters on reactivity for VVER-440 fuel (enrichment 3.6%) in condition of wet storage presented. The bounding conditions are defined. The SCALE 5.0 system is used for depletion and criticality calculation.

As new fuel designs (with higher enrichment) are introduced, spent fuel storage facilities might not fulfill criticality safety criteria when using fresh fuel approach to the criticality analyses. Since optimum moderation conditions evaluation even in wet storage systems is required by some regulatory bodies, any credit for soluble boron cannot be taken. Thus, the only suitable way to prove subcriticality of the given spent fuel system with higher enriched/burnt fuel is burnup credit implementation.
This paper outlines burnup credit implementation methodology as demonstrated on criticality evaluation of VVER-440 reactor pool at NPP Dukovany. Operational history effects, isotopic set choice, as well as computational issues (SCALE 4.4a was used both for depletion and criticality calculations) are discussed, maintaining strictly conservative approach.
Bounding approach in operational history treatment was carefully examined. Criticality evaluation using selection of (as expected) conservative values of operational parameters (specific power, fuel and moderator temperatures, boron content in moderator,..) was compared to criticality evaluation of real fuel assemblies from the NPP database. Therefore, bounding approach was justified and it was shown that it is not excessively conservative.
Presented methodology can be applied on any similar spent fuel facility. Suggestions for future research are noted (mainly end-effect evaluation and consideration of profiled fuel) and urgent need of validation of depletion codes for VVER systems is emphasized

According to the fresh fuel management procedure, fuel assemblies – after nuclear fuel delivery to the NPP fresh fuel unit – are vertically loaded into a cover intended for the delivery of fuel assemblies into the containment of the NPP reactor compartment. The cover is placed into an universal jack in the cooling and refueling pond, and then the fresh fuel assemblies are loaded into the reactor core. Based on the nuclear safety analysis carried out by the Russian Research Center “Kurchatov Institute” for contemporary WWER-1000 fuel, it has become necessary to limit the number of fuel assemblies loaded into a cover below its designed capacity (12 FA instead of 18 FA as originally designed). Such a decision leads to worse economic performances in fuel transportation. The paper considers potential ways to overcome this restriction.
Transport container TUK-6 for spent fuel assemblies was designed quite a long time ago and, as shown in this paper, the requirement on the maximally permissible neutron multiplication factor of the loaded container for individual states to be analyzed in compliance with Ukrainian regulations is not met. First of all, this concerns the container criticality analysis in optimal neutron slow-down (container filling with water-air mixture with optimal density). The paper shows potential ways for TUK-6 burnup-credit loading with the maximum number of fuel assemblies and partial container loading.

An accelerated monitoring system designed for the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice bases on the new designed “cesium detectors” is presented in the paper. Since 1999, leak tests of WWER-440 fuel assemblies are provided by special leak tightness detection system “Sipping in Pool” delivered by Framatome-anp with external heating for the precise defects determination. Although, this system seems to be very effective, the detection time of all fuel assemblies in one storage pool is too long. Therefore, a new “on-line” detection system, based on new sorbent NIFSIL for effective 134Cs and 137Cs activity was developed. Design of this detection system and its application possibility in Slovak wet interim spent fuel storage facility as well as the first results are presented.

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Overview of “realistic” alternative nuclear fuel cycles for countries – nuclear energy users is given. Results of fast reactor calculations are presented.

The first step of the investigation of the thorium fuel cycle with HELIOS 1.8 is validation of the results obtained from the code for this particular type of fuel. To complete this first task we performed calculation of the benchmark announced by IAEA in 1995. The benchmark was based on a simplified PWR model of the assembly with reduced fuel composition. This calculation was focused on a comparison of the methods and basic nuclear data.
After successful validation of the code we focused our work on calculating the PWR and VVER thorium fuel cycles. The thorium cycle begins after the first use of UO2 fuel in the reactor as separation of plutonium from the burnt fuel. Separated plutonium is mixed with thorium and used as a new nuclear fuel in the reactor. For our calculation we prepared two variants of the assembly – the first variant is a homogeneous distribution and the second one is a non-homogenised distribution of thorium fuel in the assembly. The model of non-homogenised distribution of Pu-Th fuel was designed by replacing selected rods of the classical UO2 assembly by Pu-Th rods. These selected rods are distributed symmetrically in the assembly. Other rods in the assembly remain the same as in the classical UO2 assembly.
The calculated and compared values are criticality and fuel composition as a function of burnup.

The Institute for Energy Technology is an international research institute for energy and nuclear technology in Norway. This institute operates the Halden research reactor which is used for fuel research and development in all nuclear fields.
This report is a short review of executed experiments of advanced fuels like inert matrix fuel, thorium fuel and MOX fuel which have been conducted in the last three years.

Methodology for comparison of the long term safety of nuclear fuel cycles is discussed in this paper. The first step in this process is the collection of indicators which can be used to quantify the impact of each fuel cycle. Indicators such as activity fluxes to the biosphere, annual dose for a member of critical group of people and others were calculated to show how different fuel cycles modify the performace of the disposal system. We present here safety assessment results of evolution and human intrusion scenarios. Results for evolution scenarion were calculated by software AMBER, which is able to build dynamic compartmental models to represent the migration and fate of contaminants in the environment. Second part of this paper deals with the intruder scenario, particularly radiation load of workers who analyze borehole core excavated from the deep geological repository. Radiation load was calculated by VISIPLAN 3D ALARA planning tool software. A number of calculating models was defined for each of the fuel cycles. Calculating model estimates radiation load as external irradiation of people working near active borehole core. The results of these models are presented by dosimetric maps in the neighbourhood of borehole core and tables containing dose rate during working activities. The methodology presented in the paper can be finally used for the sustainability assessment of the advanced nuclear fuel cycles based on the Multi-Criteria Analysis approach.

The main aim of the work is to give into the mutual harmony long term planning of nuclear energy (contemporary power station has supposed lifetime 70 years and will need at least ten years from the decision to full realization) with the possibility of raw materials supply during the whole lifetime. But situation is not so uniquely directed only to nuclear energy – classical sources (coal, gas and oil) are supposed to be exhausted surely in this century (coal could be regionally dependent) and specially Europe will be practically net consumer of energy raw materials.
Ways to stabilize market could be production of hydrogen and use of U238 and Th232 in the new reactors. But developments needed to influence practical industrial demand is about 30 years to be able to start to build up industrial new type nuclear power-stations and money of the order of several billion dollars. At the same time political unification has consequence on local raw material resources, which could hardly be interpreted only as the local or national ownership. Their price could sharply rise and there will be pressure to use it not only for energy production, but also as the chemical raw material. Moreover it is expected that in the half of this century the price of natural uranium will be limiting factor for the contemporary PWR reactors use and that there will be necessary to start fast breeders or use thorium fuel to cover demands. Last estimations about the oil peaking and its consequences are included.
Armenia, Lithuania and Ukraine representatives expressed will to join the project.
Common will is expressed in the presentation to try to understand commonly the future development and to find common solution to the future challenges.

There is lot of information bringing our attention to the problem of limited raw material resources. Fortunately uranium for nuclear energy is very concentrated source and that is why its transport brings no problems and could be realized from anywhere.
Second question is if overall resources are available for current nuclear energy development. Data documenting reasons for nowadays price growth are presenting and it is clearly shown that the most probable explanation is that there is gap in new uranium mines preparation and the lot of smaller mines were closed in the period of low uranium prices.
Conclusion is that there is at least for the first half of this century even for thermal reactors enough uranium. Situation could be changed if there will massive production of liquid fuel using hydrogen, produced through nuclear heating.
Public information about former military uranium resources are also included. Contemporary about one half of US nuclear power-stations is using high enriched uranium diluted with natural uranium – it is expected to continue this way up to 2012. Uranium is complicated market.

The overall objective of the HPLWR project was to assess the feasibility of a High Efficiency Light Water Reactor operating at thermodynamically supercritical region. In a once-through concept, the water enters the reactor as water and exits as high pressure steam without change of phase. An efficiency of over 40% is expected. The evaluation and improvement of the Japanese concept was carried out by eight institutions within the 5th Framework Programme of the EC. Our role was to perform neutronic transport and core diffusion calculations. As the coolant density along the axial direction shows remarkable change, coupled neutronic-thermohydraulic calculations are essential which take into account the heating of moderator in the special water rods of the hexagonal assemblies. We have prepared a parametrized diffusion cross section library of the HPLWR assembly at low burnup. The parameter range covers the cold zero power and hot full power states. The feedback parameters are calculated with the SPROD code of the Tokyo University, which was coupled with the KARATE core calculation module. Preliminary core calculations have been done to demonstrate the applicability of the code. During the HPLWR core calculations the two energy group structure was used in the coupled KARATE-SPROD code system. Because of the coolant axial density change the need for checking the applicability of the two-group structure emerged. Cross section calculations were performed with the MULTICELL deterministic transport code for the reference fuel assembly without Gd absorbers at zero burnup. The technological parameter range (coolant density, water rod density, 135Xe concentration, fuel temperature) corresponded to the nominal conditions. The multigroup cross sections were collapsed to 2, 4 and 6 group diffusion type cross sections using the criticality spectrum (B1 equations) for each parameter combination. The cross sections were used to calculate one representative assembly. The temperature and equilibrium Xe distributions were calculated with the KARATE-SPROD nodal code using 2 energy groups. The SNAP finite difference (FD) code was used to study the effect of the 2, 4 and 6 group scheme, where the cross sections were based on the frozen distributions. The results showed that the maximum deviation between the multiplication factors was about 0.2 %, so the use of the limited number of groups in this calculational scheme does not lead to considerable error.