17th Symposium of AER on VVER Reactor Physics and Reactor Safety

Control Assembly ejection in VVER-440 initiated by the loss of integrity of the Control Assemblies drive housing has been analyzed. This event causes a very rapid reactivity insertion to the core and small break LOCA which potentially could lead to rapid power increase and redistribution of heat release in the core resulting in a fuel, cladding and coolant temperature rise; primary pressure increase, radiological consequences due to loss of primary coolant and potential loss of cladding integrity and fuel disintegration (if applicable). Methodology of the analysis is based on conservative assumptions as well as on deterministic approach for selection of functioning logic of systems and equipments to maximize reactor core power and minimize power decreasing reactivity feedback. Computational analyses were performed by 3D kinetics PARCS-RELAP coupled code. VVER-440 fuel cross-section libraries, diffusion coefficients and kinetics parameters were calculated by HELOS code. In this paper analysis of accident for Hot Full Power was presented. Results of analysis show that ANPP VVER-440 reactor design meets acceptance criteria prescribed for RIA type design based accidents.

Main stages of Russian uranium fuel development regarding improvement of safety and economics of fuel load operation are presented. Intervals of possible changes in fuel cycle duration have been demonstrated for the use of current and perspective fuel. Examples of equilibrium fuel load patterns have been demonstrated and main core neutronics parameters have been presented. Problems on the use of axial blankets with reduced enrichment in VVER-1000 fuel assemblies are considered. Some results are presented regarding core neutronic characteristics of VVER-1000 at the use of regenerated uranium and uranium-plutonium fuel. Examples of equilibrium fuel cycles for the core partially loaded with MOX fuel from weapon-grade plutonium are also considered.

RUSSIAN FUEL ASSEMBLIES IMPLEMENTATION EXPERIENCE AT SUNPP-2

This paper demonstrates the recent years experience in core design and operating in the Loviisa NPP. The two units of Loviisa NPP are run with reduced core and 1500?MW power. Both units are currently in nearly equilibrium 3-batch fuel cycle. Two different fuel types are used: first unit (Loviisa-1) runs with BNFL fuel whereas TVEL fuel is used in the 2nd unit (Loviisa-2).

The paper summarizes VVER-440 second-generation fuel operation experience in improved fuel cycles using the example of Kola NPP units 3 and 4. Basic parameters of fuel assemblies, fuel rods and uranium-gadolinium fuel rods, as well as the principal neutronic parameters and burn-up achieved in fuel assemblies are presented. The paper also contains some data concerning the activity of coolant during operation.

A power uprating project actually in progress on Paks NPP units in Hungary. Nominal power of Unit 4 and 1 has already increased to 1485 MWth (500 MWe), it is the 108% of former nominal power. In order to provide economic fuel cycle on uprated power, Paks tends to introduce higher enriched fuel. The fuel has modified geometry offered by fuel vendor, the pinwise enrichment profile is slightly modified according to Paks requirements.

The paper has as an objective to make conclusions for works performed on modernization of VVER-440 fuel cycles. Possibilities and perspectives on further improvement of neutronics and operation characteristics of VVER-440 are discussed basing on additional optimizations of fuel assembly design and of fuel load patterns. Discussion is applied to reactor operation with power levels of 100 and 112% comprising equilibrium refueling regimes.

The paper touches upon the issues of NPP safety ensuring at the stage of fuel load design and operation by applying special limitations for a series of parameters, that is, design limits. Two following approaches are compared: the one used by west specialists for the PWR reactor and the Russian approach employed for the VVER reactor. The closeness of approaches is established, differences that are mainly peculiarities of terms are noted.

The paper presents the analysis of adopted safety barriers against propagation of fission product released from VVER core of active power plants..

Present paper deals with some aspects influencing significantly cost of nuclear fuel and possibilities of its usage in optimal fuel cycle technology. Our discussion is focused on the phase of fuel procurement that means financial parts of the contract as well as its technical Appendices. Typically the fuel fabrication price is taken as the main economy indicator; nevertheless also many other financial and technical features of the contract must be taken into account in order to reach the best price of electricity sold into public energy grid. Our experience from several international tenders shows that the consistent complex of commercial and technical parameters of the contract is necessary to achieve optimal economic results and prepare proper conditions for advanced fuel cycle technology. Among those essential characteristics are payment conditions and schedule and extent of vendor?s services and assistance to the operator. Very important role play also technical parameters, as safety and operational limits, influencing loading pattern quality and operating flexibility. Obviously also a level of operator?s fuel cycle technology is a crucial point that is necessary for usage of technical quality of the fuel at the power plant.
The final electricity price, produced by the plant, and uranium consumption are the only objective criteria to evaluate economic level of the fuel contract and the fuel cycle at all.

17th Symposium of AER on VVER Reactor Physics and Reactor Safety,

Joint AER WG A and WG B held their sixteenth meeting in Hrotovice (near NPP Dukovany), Czech Republic, during the period of 24-25 April 2007.
The objectives of the meeting content of presentations and future activities are shortly described in this paper.

During using of the macrocode Moby-Dick has surfaced that structure of the program has some limitations in accuracy of calculations. Many of these limitations originate in extensive using of the single precision representation of the real variables. These, along with other limitations, were removed and results of the comparison between old and new calculations are presented. Example of the calculation of the reactivity coefficients by the perturbation theory and by direct calculation is shown.

Results of thermal-mechanical behaviour 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 160 fpd and for different type of regulation. Additionally calculations were performed for campaign #7.

The paper presents new 3-dimensional neutron kinetics code for VVER type core. DiFis 2.0 is an extension of previous steady-state version of the code DiFis for transition processes.

A new cyclic iterative method with variable parameters is proposed for accelerating the outer iterations in a proposed used to calculate K[eff] in multigroup problems. The method is based on the use of special extremal polynomials that are distinct from Chebyshev polynomials and take into account the specific nature of the problem. To accelerate the convergence with respect to K[eff], the use of three Orthogonal functionals is proposed. Their values simultaneously determine the three maximal eigenvalues. The proposed method was

The paper dedicated to the test problem of pin power distribution in the fuel assembly dependent on interassembly water gap. Necessity for the solution of an offered test problem is dictated by that in process of operation of reactor VVER-1000 fuel assemblies deformations and change of distance between them were observed. That leads to local change of power release in peripheral rows of pins. In spite of the fact that in a design of last generation of fuel assemblies the scale of deformation is reducing, an estimation of possible increase of power release remains an actual problem.

The High Performance Light Water Reactor (HPLWR) belongs to the six reactor types currently being investigated within the framework of the Generation IV International Forum: the Supercritical Water Cooled Reactor ¹. The paper summarizes the activity of KFKI Atomic Energy Research Institute in the field of core design carried out within the framework of the “HPLWR Phase 2” project.

The present report covers object, conception, engineering solution of construction of modern system of high-powered reactor in-core control, including VVER-1000 (V-320) reactors.

The paper discusses investigation results of SPND signals for noises caused by fluctuations of coolant parameters. The dependence of neutron flux amplitude oscillations (local power) in SPD locations on the value and frequency of fluctuations of coolant parameters (pressure, flow rate, temperature) is analysed in the paper.

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.

Computation kernel of In-core Monitoring System (IMS) based on ?Kruiz? software is developing step by step in accordance with ?Ukrainian NPP WWER-1000 In-core Monitoring System enhancement conception? approved by Utility. By this time IMS computation kernel based on ?Kruiz? put into operation at the 5 WWER-1000 NPP units in Ukraine. In the framework of realization of ?Ukrainian NPP WWER-1000 In-core Monitoring System enhancement conception? previous activity, current situation and calculation functions evolution plans are overviewed in this report.

General overview of power distribution rebuilding algorithm, that implemented in ?Kruiz? software, is given in the report. This algorithm is based on join solution of measurements equations and neutron physics model of reactor core. Influence of some operating factors and errors of model on the power distribution rebuilding precision is estimated. Examples of the power distribution rebuilding are given using model power distribution calculated by independent codes. It is shown that the measurement accuracy gives the main contribution into the accuracy of power distribution rebuilding. Results of power distribution rebuilding algorithm verification obtained during the operation Rovno NPP unit 4 are shown.

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In year 2006 we started data processing from the Dukovany NPP operating history database that contained data from the old measurement system VK3 and the new Scorpio-VVER. The work has been done in cooperation with the reactor physicists at Dukovany NPP. Obtained data from database were compared with calculated parameters from 3D diffusion macrocode MOBYDICK.

It is important to determine the thermal power at the four VVER-440 type reactors at NPP Paks as well as at all the other reactors in the world. In these reactors the power is calculated from the difference between outlet and inlet measured temperature of cooler of primary circuit with accurate resistance thermometers. Otherwise this power is calculated on the 210 measurement of fuel assembly outlet temperature with thermocouples just above the fuels. These two calculated values – including bypass – show discrepancy. The calculated power from fuel outlet temperatures is higher than the calculated one from cooler inlet-outlet total power. Till that time we assume two reasons. One of them explains this discrepancy by imperfect mixing of cooler in head of fuels. This effect is investigated by more institutes.

In VVER-440/213 type reactors the outlet temperature of coolant is measured with thermocouples installed above 210 fuel assemblies. These temperatures are fundamental information in on-line core monitoring therefore the accuracy of the temperature measurement is very important.

Detailed knowledge of the flow and the heat transfer is very important in the case of fuel rod bundles of nuclear reactors from the design and safe operation point of view. Investigations of the thermal hydraulic processes in VVER-440 fuel assemblies are in progress at the Institute of Nuclear Techniques, Budapest University of Technology and Economics using the code ANSYS CFX 11.0 ¹. In the present work, one single subchannel was investigated in order to reach the necessary resolution of the optimal mesh at first. Models of a subchannel (triangular array, P/D=1.35) were built and comprehensive mesh sensitivity study was performed. The mesh study showed that relatively high mesh density was needed to reach the grid independent solution. Suitable mesh was chosen and turbulence model study was accomplished using the k-epsilon, SST, SSG Reynolds Stress and BSL?Reynolds Stress models. Results of the predictions were compared to measurement published by Trupp and Azad ². The best results were obtained with use of BSL Reynolds Stress model. Distributions of the Reynolds Stresses and the turbulent kinetic energy were in good agreement with Trupp and Azad?s measurement. Based on the experience of the study two further models of a 250 mm long rod bundle section were developed that contained six rods. One of these models involved a spacer grid to investigate its effects on the flow. Using these models calculations were carried out in order to investigate the mixing effects. In both cases the development of secondary flows were observed but convective mixing occurred between adjoining subchannels only in the case of the model with spacer grid.

The paper presents the last results of CFD simulation of coolant flow mixing in fuel assembly’s head. The CFD calculations were performed by VUJE, Inc. mainly for two types of fuel assemblies (FA) operated in present fuel loadings of NPP units in Slovakia (type VVER-440) – profiled 3,82 % enriched FA and Gd-II FA. The coolant outlet temperature profile in the thermocouple position was calculated for different pin-wise power distributions and FA?s burn-up. The paper describes the first step in application of CFD simulation results on measured power distributions in active core.

The main objective for the quantification of the fluid mixing in the downcomer and the lower plenum is the demonstration of the safety of the nuclear plant during non-symmetrical transients. This concerns two main topics: The risk of brittle fracture of the Reactor Pressure Vessel (RPV) due to Pressurized Thermal Shock (PTS) and the risk of core reactivity excursion during non-symmetrical transient such as Main Steam Line Breaks (MSLB) or Boron Dilution Transients (BDT). These scenarios are studied in the 1:5 scaled VVER-1000 reactor model at OKB ?Gidropress? in the framework of a TACIS project: ?Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration) at core inlet?.

The AER working group D on VVER reactor safety analysis held its 16th meeting in Paris, France during the period 08-09 May 2007. The meeting was hosted by the CEA France. It followed the final workshop on the OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark held at 07 May. Altogether 11 participants attend the meeting of the working group D, 7 from AER member organizations and 4 guests from non-member organizations. The co-ordinator of the working group, Mr. S. Kliem, served as chairman of the meeting.

The aim of the performed study is to find a reliable correlation between the measured thermocouples values at core outlet of reactor VVER-1000 with the coolant temperatures at the assemblies? outlets at these positions. Firstly, the solution is related with the correct interpretation of the measured results, and secondly with the correct modelling of the coolant flow at the location of the measuring device. Studies have been performed with the system coupled code ATHLET/BIPR-VVER, which allows to model the thermal-hydraulics and kinetics processes in VVER reactors in a 3D manner.

Recent studies showed the necessity of a detailed modelling of the core outlet region of the VVER-1000 reactor where the thermocouples are located. Solving of this problem is of primary importance for the validation of the coupled system code ATHLET/BIPR-VVER on local parameters. Therefore, special attention is paid on the reactor pressure vessel model and its nodalization schema and in particular the fluid mixing phenomena at assemblies? outlets. For this purpose additional thermal-hydraulic channels modelling the flow along the guide tubes are introduced in the reactor core.
With the new advanced modelling again the benchmark problems of Phase 1 of the CEA-NEA/OECD VVER-1000 Coolant Transient Benchmark are analysed. On the base of data comparison with the experimental measurements (Phase 2, Exercise 1) the mixing phenomena at assembly head is estimated and mixing coefficients are introduced in the thermal-hydraulic core outlet models of the coupled system code ATHLET/BIPR-VVER.

A detailed model of the primary and secondary loop of the new design of VVER-1200 NPP is being created for the coupled system code ATHLET/BIPR-VVER. On the basis of the previously gained experience, a very detailed 3D modeling of the reactor pressure vessel (RPV) and of the steam generators (SG) is being successfully applied. The nodalization schemas of these objects are chosen to be optimal ones concerning the fidelity to the real geometry and the needed CPU time. The thermal fluid objects (TFO) are modeled in ATHLET as objects of type ?pipe? most of them connected with cross flows, that allow to describe the mixing phenomena in RPV and in SGs near to reality. A pre-processor system supports automatically to prepare the complex and great number of nodalized volumes for the ATHLET input. A detailed modeling of the control and safety systems covers a wide spectrum of initiating events. Generic design data are used to model the 3D neutron-kinetics in BIPR, applying the modernized fuel assembly design for VVER-1200.

Different methods of modeling of most loaded fuel pin are presented at the work. Calculation studies are performed on example of accident related to VVER-1000 cluster rod ejection with using of spatial kinetic code DYN3D that uses nodal method to calculate distribution of neutron flux in the core. Three methods of modeling of most loaded fuel pin are considered ? flux reconstruction in fuel macrocell, pin-by-pin calculation by using of DYN3D/DERAB package and by introducing of additional ?hot channel?. Obtained results of performed studies could be used for development of calculation kinetic models during preparing of safety analysis report.

The paper presents the results of the study conducted to support planned modernization of the South Ukraine nuclear power plant. The objective of the analysis has been to develop the automated emergency control algorithm for primary to secondary LOCA accident for SUNPP VVER-1000 safety upgrading.
According to the analyses performed in the framework of SAR, given accident is the most complex for control and has the largest contribution into the CDF value. This is because of IE diagnostics is difficult, emergency control is complicated for personnel, time available for decision making and actions performing is limited with coolant inventory for make-up, probability of SDVs on affected SG non-closing after opening is high, and as a consequence containment bypass, irretrievable loss of coolant and radioactive materials release into the environment are possible.

Nowadays some Ukrainian NPPs replace their fuel storage system in the spent fuel pool with densely storage system of new design (so-called SUHT). This allows to increase significantly the number of stored fuel assemblies, that in turn results in increase of total decay power in the spent fuel pool. In this case, according to the requirements of national regulatory body and IAEA recommendations additional safety substantiations are needed. One of the main parts of such substantiation is an analysis of initiating events which lead to loss of cooling of the spent fuel pool.
One example of such a work is a safety substantiation of densely fuel storage for RNPP Units 1,2 provided by Skoda. ?Analytical Research Bureau? of SSTC NRS was taking a part in the regulatory review of this substantiation. The model of spent fuel pool for ATHLET computer code was developed to make benchmark calculations. This model allows simulating two possible cases of fuel location: (1) fuel is placed only in the densely fuel storage and (2) fuel is placed both in the upper removable and in the lower densely storages. Last feature allows checking aspects of spent fuel pool operation, which have not been estimated by Skoda.
ATHLET computer code has been selected for this task as a system thermal hydraulic code, which has a modular structure and is world wide known as an appropriate tool for safety analyses of different reactor types. Also, ATHLET allows to simulate natural convection and thermal siphon phenomena in the low parameter range (minimum temperature is 20?? and minimum pressure is 10 kPa), that is necessary for correct simulating spent fuel pool physics.
The benchmark analysis has demonstrated good agreement between results of justification analyses made by Skoda and results of comprehensive ATHLET calculations. In both cases analyses were conducted using conservative approach to the selection of initial and boundary conditions and demonstrated the acceptable level of safety.

The latest version of SCALE system (SCALE 5.1) was tested for criticality and inventory calculation for VVER-440 fuel.

Burnup of each fuel assembly conservatively uniform over assembly length and equal to average burnup of the end parts while nuclear safety analysis of the spent fuel storage taking into account burnup credit.
To avoid excessive conservatism in this approach, it is offered to use an axial conservative burnup profile for VVER-1000 spent fuel storage nuclear safety analysis.
In the report the axial conservative burnup profile is presented analyzing of spent fuel axial burnup profiles.

This paper contains a presentation and discussion of selected isotope inventory results for different types of WWER-1000 spent fuel assemblies.
Nuclide inventory calculations of spent fuel assemblies at Kozloduy NPP are routinely performed using the SCALE 4.4a computer code system. Besides the standard 17×17 ORIGEN-S library, a specific library developed at the Kozloduy NPP for each different fuel type at typical irradiation conditions is used.

Experimental validation of depletion computer codes is an ongoing need in spent fuel management. In VVER application area the lack of well?documented experimental data concerning depleted fuel is serious, being an obstacle to introduce new effective technologies and approaches in spent fuel management, e.g. burnup credit (BUC).

Spent fuel wet storage facility ISF-1 is currently used for intermediate storage of spent nuclear fuel removed from Chornobyl-1, 2 and 3.
Since commissioning of the ISF-2 dry spent fuel storage facility is significantly delayed, ISF-1 is going to be used as the main spent fuel storage facility for the Chornobyl NPP in the next few years. As ISF-1 is not capable of accommodating all SFA from ChNPP with use of the regular (design) storage scheme, compact storage of nuclear fuel in ISF-1 is under consideration.
The paper presents a comparative criticality safety analysis of regular and compact spent fuel storage in ISF-1 in operational and emergency conditions.

The study of thorium fuel cycle with plutonium as a supporting fissile material was performed under VVER-440 and PWR conditions. Our analysis was focused on the plutonium transmutation potential, amount of separated plutonium in equilibrium cycle and total amount of finally disposed plutonium. Calculations were performed by the HELIOS spectral code.

Creation fast critical molten salt reactor for burning-out minor actinides and separate long-living fission products in the closed nuclear fuel cycle (NFC) is the most perspective and actual direction. The reactor on melts salts – molten salt homogeneous reactor with the circulating fuel, working as burner and transmuter long-living radioactive nuclides in closed NFC, can serve as an effective ecological cordon from contamination of the nature long-living radiotoxic nuclides. High-flux fast critical molten-salt nuclear reactors in structure of the closed nuclear fuel cycle of the future nuclear power can effectively burning-out / transmute dangerous long-living radioactive nuclides, make radioisotopes, partially utilize plutonium and produce thermal and electric energy. Such reactor allows solving the problems constraining development of large-scale nuclear power, including fueling, minimization of radioactive waste and non-proliferation. Burning minor actinides () in MSR is capable to facilitate work solid fuel power reactors in system NP with the closed nuclear fuel cycle and to reduce transient losses at processing and fabrications fuel pins. At substantiation MSR-transmuter/burner as solvents fuel nuclides for molten-salt reactors various salts were examined, for example: LiF – BeF2; NaF – LiF – BeF2; NaF?-?LiF ; NaF?ZrF4 ; LiF?-?NaF?-KF; NaCl.

The plutonium and minor actinides are extracted from the VVER spent fuels and used in thorium fast breeder. Based on the BN800 reactor model it is investigated how the plutonium from VVER spent fuels together with thorium can be used for long term energy production. Three cases are considered:

sodium cooling, thorium is in upper and lower blanket

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

The Kh-1 power unit reactor control by using of design algorithm. Application of additional recommendations within the limits of power distribution fixed axial offset maintaining method.

Possibility of NPP operation in the electric load-follow conditions is one of the
urgent requirements established for Russian and foreign designs. NPP operating schedules in
these conditions can vary in different national power grids, though they have common
features. Daily decrease of power for 6-8 hours within the limits of the regulating range with
changing power in both directions during one-two hours is a typical requirement which, as a
rule, corresponds to the rate of external power consumers disconnection at night and their
connection by days.
The results of daily grid load-follow conditions for VVER-1000 reactor are given in
the paper. Two versions of transients are considered. In the first one the thermal change at the
reactor core inlet is described in accordance with the condition of constant pressure
maintenance in the steam generator. In the second one a combined model is simulated in
which a constant average temperature of coolant in the reactor core is maintained in the upper
area of the regulating range (from 80 up to 100 % of Nnom).
The load-follow conditions are simulated using RC code meant for three-dimensional
small-group diffusion calculation of VVER-type reactors while analyzing the concept of
automatic keeping from current offset near the target value in the range of ?5 %.

The computer simulation of ZpNPP Unit #3 (WWER-1000) Cycles 16-19 core depletion has been performed on the basis of the operational data. The changes in reactor heat rate, lead bank position and inlet temperature during the core operation have been taken into account.

The previous report ¹ on the same subject represents a brief description of procedure laid down in the code package SAPFIR _95&RC and gives calculation results of fuel cycles by operational data of VVER-440 power Units (Kola NPP, Unit No.1 and ?Dukovany? NPP, Unit No.2).

It is discussed the results of investigations on comparison of the calculated and measured FAs relative power distribution based on thermocouples data in the VVER-440 cores. All main results were obtained during 5 last years in the RRC ?Kurchatov Institute? and are based on the measurements and calculations which have been performed for the 3-d and 4-th units of the Kola NPP. Fuel loadings of various compositions were analyzed for various cycle moments. In all, about 1200 core conditions were studied.

Realistic modeling of heat transfer in fuel rods is very important for coupled computer codes, because temperature-coefficient feedback is responsible for interfaces between neutronic and thermal-hydraulic parameters of VVER reactor core.

The ZR-6 critical experiments and the validation of lattice physics codes and their cross section library are described. The parameter of interest in this analysis is the temperature reactivity coefficient. It is very important in design, control and safety, particularly in the light water reactors the VVER plants are belonging to. The agreement between calculations and measurements differs for different codes using different nuclear data. Nowadays calculations have been performed by the French code APOLLO-2 version 2.7 using the JEF-2.2 based CEA93.V7 group library. The work is a part of the NURESIM project, where KFKI AEKI undertook to develop and qualify some calculation schemes for hexagonal problems.

In conjunction with each refuelling shutdown of the reactor core, nuclear design calculations are performed to ensure that the reactor physics characteristics of the new core will be consistent with the safety limits. Prior to return to normal operation, a physics test program is required to determine if the operating characteristics of the core are consistent with the design predictions and to ensure that the core can be operated as designed. Successful completion of the physics test program is demonstrated when the test results agree with the predicted results within predetermined test criteria. Successful completion of the physics test program and successful completion of other tests which are performed after each refuelling provides assurance that the plant can be operated as designed.

Using modeled VVER scram rod drop experiments, performed at the Rostov NPP, as an example, the influence of delayed neutron parameters on the modeling results was investigated. The delayed neutron parameter values were taken from both domestic and foreign nuclear databases.
Numerical modeling was carried out on the basis of SAPFIR_95&RC_VVER program package. Parameters of delayed neutrons were acquired from ENDF/B-VI and BN?B-78 validated data files.
It was demonstrated that using delay fraction data from different databases in reactivity meters led to significantly different reactivity results.

The method activation researches parameters of neutron energy spectral are developed in RRC KI. The method use multi component detectors, which were named ?Unified Composite Detectors? (UKD). Methodical basis of the measurements is Westcott formalism for representation of a spectrum in thermal and epithermal areas of energies [1, 2]. Parameters of the spectrum are: Ft – thermal flux, fepi – parameter epithermal flux and ?n – temperature of neutron gas. The detector includes nuclides – targets 164Dy, 55Mn, 197Au, 186W, 81Br. They are used for determination of absolute values of DNF of thermal (Ft) and parameter (fepi) epithermal neutrons: nuclides 164Dy, 55Mn – for thermal area and nuclides 197Au, 186W, 81Br – for epithermal areas – (up to 120 eV). Detectors with nuclide 115In are used for estimations of DNF in transitional area of the spectrum.

The documentation, required for permission to carry out changes influencing nuclear safety, has to cover according to Czech Atomic law:

17th Symposium of AER on VVER Reactor Physics and Reactor Safety,