REACTOR DOSIMETRY: ACCURATE DETERMINATION AND BENCHMARKING OF RADIATON FIELD PARAMETER, RELEVANT FOR PRESSURE VESSEL MONITORING (REDOS)

Transcription

1 AMES Ageing Materials European Strategy REACTOR DOSIMETRY: ACCURATE DETERMINATION AND BENCHMARKING OF RADIATON FIELD PARAMETER, RELEVANT FOR PRESSURE VESSEL MONITORING (REDOS) REDOS Final Report AMES Report n.17

2 NOTE FROM THE EDITOR The whole set of AMES Reports: Myrddin Davies, Debarberis L., Lyssakov V., Gillemot F., Sevini F., Ballesteros A. "Nuclear Power Plant Life Management in Some European Countries", Davies AMES Report EUR EN, 2002 Ageing Materials Evaluation and Studies by Non-destructive Techniques, EUR 19675EN - December 2000 [19 MB] A.Ballesteros et al., Proceedings of the AMES Workshop on Reactor Pressure Vessel Life Predictions, EUR 19638EN - November 1999 A. Ballesteros, H.Ait Abderrahim, L.Debarberis, T.Lewis, C.Sciolla, T.Seren, M.Valo, W.Voorbraak, B.Acosta, Conversion Table of Material Damage Indexation for all Different European Reactor Types, EUR 18693bEN - February 1999 Antonio Ballesteros, Wim Voorbraak, Luigi Debarberis, Dosimetry and Irradiation Programmes of AMES European Network, EUR 17744EN - December 1997 CEA, TECNATOM, VTT, Effect of Irradiation on Water Reactor Internals, EUR 17694EN - June 1997 L.M. Davies, A Comparison of Western and Eastern Nuclear Reactor Pressure Vessel Steels, EUR 17327EN - April 1997 Pierre Petrequin, A Review of Formulas for Predicting Irradiation Embrittlement of Reactor Vessels Materials, EUR 16455EN - November 1996 Juan Bros, Carlos Cueto-Felgueroso, Antonio Ballesteros, Dosimetry and Neutron Transport Methods for Reactor Pressure Vessels, EUR 16470EN - November 1996 Luigi Debarberis, Lay Tjoa, AMES Reference Laboratory JRC-IAM/ECN Petten, EUR 16409EN - June 1996 J. Föhl, Survey of Existing Planned and Required Standards, EUR 16316EN - December 1995 R.Gerard, Survey of National Regulatory Requirements, EUR 16305EN - June 1995 Kim Wallin, Comparison of the Scientific Basis of Russian and European Approaches for Evaluating Irradiation Effects in Reactor Pressure Vessels, EUR 16279EN - February 1995 Reijo Pelli, Kari Törrönen, State-of-the-Art Review on Thermal Annealing, EUR 16278EN - December 1994 Tapio Planman, Reijo Pelli, Kari Törrönen, Irradiation Embrittlement Mitigation, EUR 16072EN - September 1994 can be downloaded at: Filippo SEVINI (AMES Network manager) Aug. 30 th,

3 Table of Contents 1 SUMMARY ACKNOWLEDGEMENTS INTRODUCTION OBJECTIVES WORK PROGRAMME WP1. DATA COLLECTING THE LIGHT WATER EXPERIMENTAL REACTOR LR WWER-1000 MOCK-UP DESCRIPTION AND EXPERIMENTAL DATA WWER-440 MOCK-UPS DESCRIPTIONS WWER-440 Mock-up WWER-440 Mock-up WWER-440 EXPERIMENTAL DATA Measurements and evaluations Evaluated experimental data REVIEW OF NPP AVAILABLE EXPERIMENTAL DATA WP2. EXPERIMENTAL ACTIVITIES. PHOTON SPECTRA MEASUREMENTS OVER THE RPV SIMULATOR IN THE LR0 REACTOR INTRODUCTION TO THE PHOTON SPECTRA MEASUREMENTS SPACE-ENERGY DISTRIBUTION MEASUREMENTS OF NEUTRON AND PHOTONS RESULTS OF THE EXPERIMENTAL WORK OTHER MEASUREMENTS CONCLUSIONS OF THE EXPERIMENTAL WORK WP3. BENCHMARKING. TRANSPORT CALCULATIONS AND ANALYSIS OF MEASURED AND CALCULATED DATA INTRODUCTION TO THE DATA ANALYSIS METHODOLOGY ANALYSIS OF CALCULATION RESULTS FOR THE WWER-1000 MOCK-UP Points of measurements and calculations Comparison of the calculation results Comparison of calculated and experimental results Influence of the detector on the experimental results ANALYSIS OF CALCULATION RESULTS FOR THE WWER-440 MOCK-UP 1 AND THE WWER-440 MOCK-UP Points of measurements and calculations Comparison of calculation results of participants Comparison of calculated and experimental results EVALUATION OF UNCERTAINTIES AND ANALYSIS OF RESULTS Sensitivities of results to different basic nuclear data libraries CONCLUSIONS OF THE BENCHMARK EXERCISE WP4. RPV ATTENUATION IN POWER REACTORS ATTENUATION THROUGH THE RPV WALL Introduction to RPV Wall Attenuation Different Approaches

4 9.1.3 RPV Attenuation in WWERs Conclusions on RPV Wall Attenuation APPLICATION OF THE LR0-WWER MOCK-UPS RESULTS TO POWER REACTORS Introduction to the conformity study Comparison of the Mock-ups and NPP results Conclusions of the conformity study GENERAL CONCLUSIONS & RECOMMENDATIONS REFERENCES

5 Reactor Dosimetry: Accurate Determination and Benchmarking of Radiation Field Parameters, relevant for Pressure Vessel Monitoring (REDOS) A. Ballesteros 1, X. Jardí 1, B. Böhmer 2, J. Konheiser 2, B. Ošmera 3, J. Kyncl 3, A. Keresztúri 4, G. Hordosy 4, K. Ilieva 5, S. Belousov 5, D. Kirilova 5, M. Mitev 5, B. Petrov 5, B. Acosta 6, V. Smutný 7, E. Polke 8, S. Zaritsky 9, C. Töre 10, P. Ortego 10 Shared Cost Action / FIKS-CT ) Tecnatom, Madrid, Spain 2) FZ Rossendorf, Dresden, Germany 3) Nuclear Research Institute, Řež, Czech Republic 4) AEKI, Budapest, Hungary 5) INRNE, Sofia, Bulgaria 6) JRC-IE, Petten, The Netherlands 7) Škoda, Plzen, Czech Republic 8) Framatome ANP GmbH, Erlangen, Germany 9) Kurchatov Institute, Moscow, Russia 10) SEA, Madrid, Spain 1 SUMMARY The REDOS project aims to improve reactor dosimetry for radiation embrittlement monitoring of the reactor pressure vessel (RPV) steels. Benchmarking, as well as combined experimental and computational techniques, have been used. Specific objectives were the improvement of the neutron-gamma calculation methodologies through the LR-0 engineering benchmarks for WWER-1000 and WWER-440 reactor types, and the accurate determination of radiation field parameters in the vicinity and over the thickness of the RPV. A review of the available experimental reactor dosimetry data of Kozloduy Units 1, 4, 5 was performed, and attenuation coefficients through the vessel wall were calculated. Existing data for the WWER-440 and WWER-1000 Mock-ups were reviewed and the preparatory work for subsequent measurement and experimental data analysis was carried out. Measurements of the space-energy distribution of the mixed neutron-gamma field in the WWER-1000 model over the RPV simulator were performed. The most important improvements in the experimental techniques used were the multiparameter spectrometer and a new low noise precise monitoring system in the LR-0 research reactor, developed for this type of measurements. 4

6 For the WWER-1000 Mock-up and the two WWER-440 Mock-ups and for all positions, where measurements had been performed, neutron-gamma transport calculations were carried out independently by the participants using deterministic and/or stochastic codes and associated nuclear data libraries, mostly based on ENDF/B- VI. The calculated neutron and gamma flux integrals, DPA-rates and spectra were compared with each other and with experimental values. Seven institutions from 5 countries (Bulgaria, Czech Republic, Germany, Hungary and Spain) delivered calculation results for the WWER-1000 Mock-up, six institutions from 4 countries performed calculations for one or both WWER-440 Mock-ups. More than eight different calculational schemes were used, covering the most important methods used for pressure vessel dosimetry and shielding. The codes used were the stochastic codes MCNP and TRAMO and the deterministic codes ANISN/DORT and TORT. They were associated with different data libraries and data preparation schemes. The BUGLE 47n/20g group structure was used for comparisons of calculated spectra as well as for comparison with experimental spectra. Results of RPV attenuation calculations for WWER-440s and WWER-1000s were obtained. A comparison between Western and Eastern attenuation determination approaches was carried out. The neutron/fluence (E>0.5 MeV) wall attenuation for WWER-440 and WWER-1000 RPVs is slower than the dpa attenuation. This means that fluence above 0.5 MeV is more conservative approach than the use of dpa. Another task performed was the application/extrapolation of the WWER mock-ups results to power reactors. The attenuation through the RPV of the neutron flux/fluence with energy above 0.5 MeV was determined. The relative difference value does not exceed 10% for WWER-440. The relative difference value does not exceed 10% for WWER-1000 too, except at the position behind the RPV wall. 2 ACKNOWLEDGEMENTS This work was executed under the multi-partners research contract REDOS N FIKS- CT , co-financed by the European Commission under the Euratom Specific Nuclear Fission Safety Programme The project partners wish to acknowledge Dr. S. Casalta and Dr. G. Van Goethem from European Commission DG Research for the excellent project follow-up and guidance. Thanks also to Mr. D. Bekriev, Ms. V. Tzocheva, Ms. R. Ivanova from Kozloduy NPP for the data supplied and their interest in the REDOS results. Scientific information and engineering know-how has been made available in the hope that this will lead to improved pressure vessel safety and integrity. 5

7 3 INTRODUCTION Radiation-induced damage to the RPV steel is mainly due to high-energy neutrons. The assessment of the RPV fast-neutron fluence is therefore the most important aspect for the evaluation of the vessel steel embrittlement. For all power plants, specific surveillance programs have been instituted in order to monitor the irradiation effects on RPV material [1]. These programs typically involve irradiation of surveillance capsules inside and outside the vessel, capsules which are equipped with neutron dosimeters and samples of material representing the vessel material. The neutron fluence is evaluated combining dosimetry results and transport calculations and is, then, used in some damage model to deduce the damage in the material. It is apparent the close relationship between experimental and computational results. Evaluation of neutron spectra is necessary to translate the results of the dosimetry measurements into fluence data. On the other hand, the measurements are used to validate the calculation results. Also, the calculated spectra can be adjusted with measurements via least-square fitting. Results of calculations alone are subject to too many uncertainties, related for example to geometrical and compositional variables, nuclear constants characteristics, modeling assumptions, mathematical approximations, fine energy structure in the neutron spectrum, etc. The problems related to the fine energy-group structure in the spectrum, for instance, due to resonance scattering in iron, become more significant with increasing penetration in the RPV. This can lead to relevant errors if an energy-group structure of inadequate resolution is used for the calculations. The existence of these uncertainties requires, at any rate, that the calculations be benchmarked to measured results, both from controlled benchmarks (that is, clean experimental configurations designed specifically for RPV situations) [2] and plant-specific measurements. On the other hand, also measurements are subject to uncertainties. In general, given the complexity of the physical phenomena involved on the one side, and, on the other side, the variety of measurement and computational methods currently used to address the same phenomena, further standardisation and discipline is felt necessary to increase data comparability [3,4]. An accurate calculation of the neutron fluence and fluence rate is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposures parameter values in the reactor pressure vessel. Exposure parameters values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements. In recent years, nuclear utilities have become interested in extending the operating life of reactors beyond the originally licensed 30 to 40 years. The longer operating time 6

8 means higher fluences levels and/or smaller safety margins. This requires further reduction of the uncertainties of the fluence estimation procedure. Neutrons induce damage in structural materials principally through hadronic interactions of neutron scattering with, and capture by, the nucleus. This results in atomic displacement and generation of hydrogen and helium gas. Gamma rays cannot undergo such reactions. They interact by the weaker electromagnetic interaction and effect little momentum transfer. They cause very little direct displacement or gas production. However, they can energize electrons which in turn have sufficient mass to induce atomic displacement, though again by electronic interaction. The today s procedure for predicting the degree of the reactor pressure vessel material degradation takes into account the neutron irradiation only. Recent results [5,6] of material research have shown that the gamma irradiation may substantially affect the material embrittlement. Despite improvements in the calculation of neutron field parameters with corrected cross section values, remarkable discrepancies exist between calculated and measured values. The calculation accuracy of gamma field parameters in the pressure vessel environment has been evaluated to be on an essentially lower level than that of the neutron field parameters. 4 OBJECTIVES Because the diameter of WWER type RPVs was limited by the requirement for its railway transport to the nuclear power plant (NPP), the neutron exposure is relatively high and reactor dosimetry plays a very important role in WWER plant lifetime management. The space-energy distribution (differential spectra) in the vicinity of the WWER pressure vessel, which has been obtained as a result of the REDOS project, provides information in an extended range in comparison with spectra data commonly reported. Particular objectives of the REDOS project are: Improvement of RPV monitoring. Improvement of the neutron-gamma calculation methodologies through the LR-0 engineering benchmark. Accurate determination of radiation field parameters in the vicinity and over the thickness of the RPV. The project is focused on WWER reactor type, but the results are also of interest for western PWRs. 7

9 5 WORK PROGRAMME The project started up with a review of existing data, namely with the WWER-440 and WWER-1000 engineering benchmark experimental data, and NPP data relevant for attenuation coefficients through the pressure vessel wall. The experimental activity (WP2) was focused on gamma-ray spectra measurement and extended neutron spectra measurements in a mock-up in the LR-0 reactor to create a three dimensional benchmark in the vicinity of a RPV simulator (WWER-1000 engineering benchmark). Neutron and gamma transport calculations were planned for the WP3. The analysis of the benchmark data obtained with experimental and calculational methods was carried out jointly by the project partners. The results of WP4 provided accurate information on the neutron-gamma exposure parameters through the thickness of the RPV, where important changes in neutrongamma spectrum are expected. In summary, the project was divided in four work-packages. Namely, Work-package 1 Work-package 2 Work-package 3 Work-package 4 Review of available experimental data. Data collecting. Experimental measurements in the WWER-1000 Mock-up. Neutron and gamma transport calculations. Analysis of the calculated and measured data. Determination of radiation field parameters in the vicinity of and over the thickness of the RPV. Study of the applicability of the mock-up results to power reactors. Detailed information on the project results is given in references [7] to [15]. 8

10 6 WP1. DATA COLLECTING A review of available experimental data was carried out in WP1. Full-scale mock-ups of WWER-1000 and WWER-440 reactors were developed on the LR-0 reactor in NRI Řež (Czech Republic) in collaboration with Škoda NM (Plzeń, Czech Republic) and RRC Kurchatov Institute (Moscow, Russia) for research in RPV dosimetry field. These mock-ups included an appropriate sector of WWER core and full-scale (in radial direction) simulators of baffle, barrel, downcomer, pressure vessel and biological shielding. Several measurements and calculations were performed for these mock-ups. Besides that, experimental data from Kozloduy NPP (Units 1, 4 and 5) was collected for later use in WP4. In short, this work package constitutes the basis to develop the subsequent work packages. 6.1 THE LIGHT WATER EXPERIMENTAL REACTOR LR-0 The LR-0 reactor was designed for research of neutron-physical parameters of pressurized water reactors of WWER type. The reactor tank is placed in a big concrete room and measurements can be performed also outside the tank (see Figure 1). Crane Loading mast Rotating cover Moderator level meter Reactor tank Absorption cluster Horizontal channels Supporting structure Moderator inlet tube Fuel assembly Supporting plate Safety valve Moderator outlet tube Fig. 1. Reactor LR-0 general view 9

11 Suitable geometrical conditions and flexible technical arrangements of the LR-0 enabled to construct full-scale physical models of an appropriate sector of WWER type reactor in radial direction, from the core to the biological shielding. The simulators of the reactor internals (baffle, barrel) were located inside the LR-0 tank and the pressure vessel and biological shielding simulators were located outside the tank but inside the LR-0 shielding room. 6.2 WWER-1000 MOCK-UP DESCRIPTION AND EXPERIMENTAL DATA The WWER-1000 pressure vessel simulator consisted of four 5 cm thick steel layers, which can be successively moved in radial direction in order to form an air gap layer about of 65 mm width for the spectrometric measurements over the pressure vessel thickness. This gap was added to an appropriate air gap into a steel displacing tank and placed in the space between the barrel simulator and the LR-0 tank. The purpose was simulating the downcomer water density reduction, which takes place in the WWER power reactor operational conditions. The geometry and material composition of WWER-1000 Mock-up are described in [7] and Figure 2 shows a general view of the LR-0 reactor and the Mock-up. The core of WWER-1000 Mock-up was assembled on an special supplementary support plate. Fig. 2. WWER-1000 Mock-up The core power distribution (fission rate) was measured by means of the gamma scanning of the irradiated fuel pins. The radial fission rate distribution was determined 10

12 by the measurements of the pins activity at the centre of pin length in symmetrical positions (with respect to the mock-up axis of symmetry) with two independent measuring devices. The four measured values were averaged (as independent measurements) and the measured uncertainties were evaluated. The axial distribution was measured with 2 cm step using a collimator. The measured axial distribution was also averaged, separately for core boundary and out-of-boundary fuel elements. The axial extrapolated length was evaluated for each distribution. These measured data can be used for power distribution calculation benchmarking. Core calculations (critical parameters, pin-by-pin power distribution) were carried out using MOBY-DICK diffusion code [16], (Czech WWER-440 standard code) supported by WIMS-D4 code [17] for input cross-section data. The neutron spectrum was also measured in ten positions in the central plane of the Mock-up along its axis. Neutron spectra were measured with proton recoil spectrometer consisting of two stilbene scintillators and the set of 3 spherical proportional counters with different hydrogen gas filling. The measured spectra were presented in the SAILOR group format together with the uncertainties. In order to facilitate the comparison of the calculated and measured data in mock-ups and in power reactors, the integral fluxes and space-energy indices (ratios of different integral fluxes in the measuring points) were also evaluated. Spectral indices and measured attenuation factors are shown in reference [7]. The spectral indices are defined for a given spectrum as the ratio of the integral fluxes above different neutron energy thresholds, and the attenuation factors are defined as the ratios of integral fluxes above threshold energy in different points of the Mock-up. These indices represent attenuation of the integral fluxes by the water and iron layers between the barrel and the RPV. The experimental space-energy indices have been evaluated combining the measurements with neutron spectrometer and the neutron flux density power monitoring system. The long-term stability (within weeks) of the monitoring system was 5% and the short time stability was 2% (within 1-2 days). Uncertainties evaluation took into account energy scale calibration, energy resolution of the detectors, statistics, stability and noise properties of the measuring devices and the power monitoring uncertainty. 6.3 WWER-440 MOCK-UPS DESCRIPTIONS The WWER-440 Mock-ups were assembled in the LR-0 experimental reactor in NRI just like the WWER-1000 Mock-up. The basic idea of the Mock-up was the full-scale simulation (in ~ 60º geometry sector) of the radial - azimuthal geometry and material composition of the WWER-440 reactor. This geometry sector was modeled from the 11

13 core periphery area to the biological shielding. The core, basket and barrel simulators were located inside the LR-0 tank, and the RPV and biological shielding ones outside the tank. The Mock-up was not only azimuthally but also axially shortened due to the 1.25 m LR-0 fuel element active length. Two WWER-440 Mock-ups are presented under REDOS project in references [10] and [11]. The first one -with an standard core loading pattern-, and the second -with low leakage core-, i.e. with the dummy steel assembly simulators at the core periphery. The drawings for this Mock-up description were prepared using the special modules of the Monte Carlo code package MCU (RRC Kurchatov Institute) assigned for input geometry definition and description including the checking and graphical output WWER-440 Mock-up 1 The WWER-440 Mock-up No.1 description is presented in reference [10] and it can be seen in Figure 3. Detailed pictures of the mock-up geometry can also be found in reference [10]. Fig. 3. WWER-440 Mock-up No 1 Because of the lack of WWER-440 fuel assemblies in the LR-0 fuel inventory, WWER assemblies were used. To simulate the WWER-440 core periphery, three special assemblies (I, II, and III) were designed and manufactured including the 8 mm thick stainless steel WWER-440 core blanket simulator. For simulation of the water density reduction, which takes place in the WWER power reactor operational conditions, a steel displacing tank containing an appropriate air gap was placed in the space between the barrel simulator and the LR-0 tank. 12

14 Just like in the WWER-1000 Mock-up, the pressure vessel simulator consists of three 5 cm thick steel layers, which can be successively moved radially in order to form an additional air gap about of 65 mm width for the spectrometric measurements over the pressure vessel thickness. The positions of the spacing grids are given by the vertical coordinates of their centres related to the lower end of the uranium in the fuel elements. In the peripheral assemblies I, II and III of Mock-up No. 1, five non-standard grids - plates of 3 mm thickness with holes - were used. The mass of these spacing grids corresponded to the mass of the WWER-1000 type reactor grids. The upper level of water in the Mock-up No.1 coincides with upper level of the fuel WWER-440 Mock-up 2 The WWER-440 Mock-up No.2 description is presented in reference [11 and it can be seen in Figure 4. Detailed pictures of the Mock-up No.2 geometry can also be found in reference [11]. Fig. 4. WWER-440 Mock-up No 2 Six models of steel shielding assemblies were located on the periphery of Mock-up No.2 core. Only one special spacing grid (plate of 3 mm thickness with holes) per assembly was used in the peripheral assemblies I, II and III. Two special spacing grids (with the same mass as the spacing grids of the WWER-1000 Mock-up) were used in assembly

15 Special attention was paid to the background due to albedo of the core leakage neutrons at the measuring points of Mock-up Nº2. The 25 cm thick water axial reflector effectively suppresses the background in the distant measuring points far from the core. The critical height with this axial reflector used to be reached by adjustment of the boron acid concentration in water moderator during the start-up operation. Source calculations (relative core power distribution) were performed for both WWER- 440 Mock-ups with the module BORORO of MOBY-DICK computer system [16] in two groups finite difference approximation using Borresen modification. The computer path can be created using appropriate modules for 2D and 3D calculations of two types of grids, which consist of: Triangles (coarse mesh calculations by subdivision of fuel assembly to 6k2 elements). Hexagons (pin-to-pin calculations). The module BORORO is designed for the calculations of zero power core with fresh (unirradiated) fuel. In the library of two group cross sections only the dependences on global parameters (boron acid concentration, fuel and moderator temperature) and spectral index are included. The calculations were performed in a 2D pin-to-pin approximation with an effective axial leakage and Mock-ups cores were supposed to be symmetrical, i.e the nonsymmetrical control cluster was neglected. Diffusion cross-sections library was created for this purpose using WIMS-D4 code [17] excluding cross sections for dummy assemblies, where WIMS8 [18] was employed. It contains two groups diffusion cross sections for coarse and pin-to-pin calculations of WWER-1000 core type for temperature 20 ºC and three boron acid concentrations (0, 4, 8 g/kg). The cross sections are parameterized in dependence on boron content. Also the reflector cross sections are included in library (for water-steel mixture, for water and for steel). Pin-to-pin relative power calculations for Mock-ups No.1 and 2 are shown in references [10] and [11]. Relative pin powers are normalized to the mean value. The calculations were performed in the left half of the core neglecting its asymmetry due to cluster in the right half. 14

16 6.4 WWER-440 EXPERIMENTAL DATA References [10] and [11] describe completely the geometry and composition of the WWER Mock-ups, including the following information: General and detailed views of the Mock-ups. Material and enrichment specification. Loading scheme of each assembly. Core relative power distribution. In addition to this, several experiments have been performed in WWER-440 Mock-ups to support the reactor pressure vessel (RPV) dosimetry methodology qualification Measurements and evaluations The differential neutron energy spectra were measured at the crucial positions of the WWER-440, i.e. at the barrel outer surface (surveillance specimen position), inner and outer surfaces of the RPV, and inside the RPV (1/3 of its thickness). Measurement points are shown in Figures 1 and 2 of reference [9]. The spectra measurements were carried out with a proton recoil spectrometer consisting of a stilbene scintillator and a set of spherical, hydrogen filled proportional chambers with different gas pressure. The neutron spectra were measured in the horizontal central plane of the core symmetry along the Mock-up symmetry axis at selected points from the core to the biological shield simulator in the energy region from ~ 0.1 to 10 MeV. The differential neutron energy spectra measured in the Mock-up provide a good test of the calculational models and data libraries, and provide more information than a set of reaction rates usually measured in many benchmarks Evaluated experimental data Experimental results are presented in [9] in the following formats: Group neutron flux densities in the BUGLE47n/20g group energy structure, normalized separately in each measuring point in the energy interval MeV to compare the calculated and measured shapes of the spectra. Integral neutron flux densities above energies 0.l, 0.5, 1.0, 2.0, 3.0 and 5.0 MeV in the same normalization. Space - energy indices, i.e. attenuation coefficients for above mentioned integral neutron fluxes between points 2 and 3, 3 and 4, 3 and 6. Spectral index, i.e. ratio of integral neutron flux densities above 0.5 or 3.0 MeV. 15

17 6.5 REVIEW OF NPP AVAILABLE EXPERIMENTAL DATA Collecting of the NPP experimental data was performed for later use in WP4, specially 54 Mn activity measurement from scraps taken out from the inner wall and activity of the ex-vessel detectors (iron, copper, niobium) placed in the air cavity behind the vessel of Unit 1 (WWER-440/230) of Kozloduy NPP [8]. Geometry description and material specification have been reported for Unit 1 (WWER-440), Unit 4 (WWER-440) and Unit 5 (WWER 1000) of Kozloduy NPP. Measured specific activities, reaction rates and effective activation cross-sections of the ex-vessel detectors, irradiated in the vicinity of the reactor vessel, have been reported for: 18th cycle of Unit 1. 13th and 14th cycles of Unit 4. 5th and 6th cycles of Unit 5. The source have been presented by the loading schemes, operational regime and distribution of the assembly types, assembly relative powers, and burnup at the beginning and the end of the cycles. The ASYNT method [19], based on the three-dimensional adjoint solution of neutron transport equation, was applied for calculational determination of detector activity, reaction rate, and effective activation cross-section. For both types of reactors, WWER-440 and WWER-1000, two problem-oriented libraries BGL440 and BGL1000 [31] have been used. The libraries have been created by collapsing the fine-group library VITAMIN-B6 (199 neutron and 42 gamma groups) to 67-group structure (47 neutron and 20 gamma groups). The libraries consider the features (detailed 1D geometry and material compositions) of the appropriate reactor and contain upscattering data for the five thermal-energy groups. The software package used for neutron fluence calculation is presented by the flowchart in Figure 5. Comparison of measured and calculated results was performed for cycles: of Unit of Unit or Unit 5. 16

18 NESSEL-4 Neutron Macro XS DATA History DATA, Decay Constants Hexagonal SOURCES (neutron, activation, damages) DOSRC Loading/ History DATA PYTHIA TRAPEZ Assembly wise XS Libraries BGL440, BGL1000, Detectors' XS GIP FLEX DERAB Pin wise Assembly/ Pin wise Power Distribution Geometry & Material DATA (R,θ,Z)-model for DORT/ TORT Checking Body Model, (R,θ,Z)- meshes (R,θ,Z)-source (DORT/ TORT) Material XS DATA (DORT/ TORT) TORT DORT ASYNT 3D + Solution RESULTS Fluence, Activity, Damages Fig. 5. Flowchart of the Software Package for Neutron Fluence Responses Calculation 17

19 7 WP2. EXPERIMENTAL ACTIVITIES. PHOTON SPECTRA MEASUREMENTS OVER THE RPV SIMULATOR IN THE LR0 REACTOR 7.1 INTRODUCTION TO THE PHOTON SPECTRA MEASUREMENTS The multifactor dosimetry of RPV takes into account not only the fast neutron fluence. To increase the precision of the description of the mixed photon and neutron radiation field, and eliminate the probable systematic biases, a careful benchmarking should be carried out. In the frame of the REDOS work-package 2, the photon spectra were measured over the RPV thickness. The multiparameter analysis enabled simultaneous measurement of the fast neutron and photon spectra. All measurements were performed in the central plane of the benchmark and the azimuthal distributions were measured in all points in the range 0 15 degree with the zero angle at the benchmark axis. The results were obtained with an stilbene scintillator (cylindrical crystal with dimensions 4.5 x 4.5 cm). The reactor power was monitored with a low noise monitoring system and the absolute values were re-evaluated to one pulse of fixed detector. 7.2 SPACE-ENERGY DISTRIBUTION MEASUREMENTS OF NEUTRON AND PHOTONS The benchmark was operated with the 25 cm thick water axial reflector (in the top part, the bottom reflector is thicker) in order to suppress the background (albedo) of neutrons at the measuring points in the RPV simulator. The moderator contained 4.6 ± 0.1 g of boron acid per litter and the power was controlled with two symmetrically positioned clusters with three absorbing pins. A low noise monitoring system, independent of the operational one, was developed in the LR-0 laboratory. It consisted of two measuring channel with fission chambers (cylindrical proportional chambers). One was fixed and one movable to cover 4 5 orders in neutron flux for the deep penetration measurement. The preamplifiers were directly connected to the chamber, and a high voltage (HV) was created with batteries, so that the possible source of the high frequency disturbances from the electrical grid was avoided during the measurements. The HV supply was checked continuously, the batteries were recharged during not operational time. The neutron flux density was recorded every second, the measuring files were evaluated with semi empirical procedures (moveable mean value and Poison distribution tests). The power was kept constant during the measurements, averaged discrepancies from the chosen power level were displayed in the information screen of the staff. The monitors were placed out of the core in dry channels fixed in the LR-0 tank; the movable one was driven with a step-wise (2 mm) motor like the control clusters. The position of the moveable monitor was always recalibrated against the 18

20 fixed one using an appropriate power level. The long term stability and reproducibility of the monitoring system was better than 2 %. The spectra were measured in the central plane of the core with remotely controlled mapping devices enabled change the azimuthal position of the detector in the range from 5 to 15 degrees, with zero at the benchmark axis. The RPV simulator consisted of four 5 cm thick steel layers. Moving each of them radially step by step, an air gap for the spectrometer probe can be created; the azimuthal mapping device was available in all measuring points (see Figure 6), except in the point 2. Figure 6: The Azimuthal Mapping Device 7.3 RESULTS OF THE EXPERIMENTAL WORK The photon and fast neutron spectra were measured with the two-parametric spectrometer [20] using a stilbene scintilator of the cylindrical shape, diameter 4.5 cm and height 4.5 cm. The results were obtained by averaging several measurements in each point. The relative uncertainties were the standard deviations obtained during averaging. The fine group spectra were determined in the differential format, i. e. the dimension is [1/cm 2 smev], per one pulse of the monitor. The fine group spectra were collapsed into the BUGLE energy group format (Tables 1-3). The differential spectrum in 1/4 of RPV in the lethargy scale (Figure 7) demonstrates the energy resolution capability. The measurements in points 3 7 and in point 2 were carried out during two inconsistent series of measurements. There were inconsistent uncertainties of type B due to the technical reconstruction after the flood occurred in Prague in

21 Таble 1 : BUGLE Photon Spectrum in Points 3 and 4 E low Flux density POINT 3 POINT 4 Relative uncertainty [%] Flux density Relative uncertainty [%] Table 2: BUGLE Photon Spectrum in Points 5 and 6 E low Flux density POINT 5 POINT 6 Relative uncertainty [%] Flux density Relative uncertainty [%]

22 Table 3: BUGLE Photon Spectrum in Points 7 and 2 E low Flux density POINT 7 POINT 2 Relative uncertainty [%] Flux density Relative uncertainty [%] E ϕ E, MeV Figure 7: Photon spectrum in one quarter of RPV simulator 21

23 The ratio of the photon flux above 1 MeV and the fast neutron flux is presented in Table 4. The standard deviation in the point 3 was rather greater than in other points, because one measurement in each azimuthal distribution point was included in averaging; the same procedure was applied in the points 4 7. The azimuthal distribution of the neutron flux above 1 MeV in the point 3 is given in Table 5. The azimuthal distribution of the integral neutron flux above 1 MeV in the point 3 is measurable, nevertheless the standard deviations are relatively high (8-12%) comparing it with the measured effect. In point 4 the fast neutron flux distribution depended on the azimuthal angle weekly (the value was at the level of measuring uncertainty). In points 5-7 the azimuthal distribution was totally flat. From two to four measurements were performed in every azimuthal point of measurement, moving the detector from 5 o to 15 o and back. All the values were averaged. Table 4: Ratio of Photon Integral Flux above 1 MeV to the Neutron flux in Points 2 7 Point φ (γ >1 MeV) / φ (n >1 MeV) Standard deviation Table 5 Relative Azimuthal Distribution of Neutron Flux above 1 MeV in Point 3 Point of Measurement Neutron Flux with Standard Deviation Angle [ o ] 3/0 1 ± / ± / ± / ± / ± / ± / ±

24 7.4 OTHER MEASUREMENTS The thermal flux was measured in points 0, 3, 4 with 3 He chamber [20]. The chamber was calibrated in the channel 0 where the thermal flux was measured with Au foil. The bare and Cd covered Au foils were irradiated in the channel (point) 0 at the well known monitored power. The conventional thermal flux (in the peak of the Maxwell Distribution) was evaluated in accordance with a known standard procedure [21]. The fluxes published in this reference are the integrals over the Maxwell Distribution at room temperature. This procedure involves many uncertainties of type B ( systematic ), mainly due to the methodology. On the other hand, the guess of the precision of the thermal flux calculation with the deterministic codes and the BUGLE library maybe troublesome. 7.5 CONCLUSIONS OF THE EXPERIMENTAL WORK The azimuthal gradient of the fast neutron flux density was found only in the point 3 (the neutron and photon spectra agreed in all angles in the frame of experimental uncertainties) relating to the modeled part of the WWER The change of the neutron flux density over the RPV thickness was characterized by the neutron spectra measured along the model axis in points 3 7. The gamma fluxes depend on the thickness in the RPV (respective on the thickness of the water and steel layers between the core and the measuring point). The iron lines (prompt gammas) dominated in the photon spectra, except in the points 2 and 3. The ratios of the measured fast neutron and gamma fluxes were reliable. In point 2 the gamma background from the H (n, gamma) in the detector took place. 23

25 8 WP3. BENCHMARKING. TRANSPORT CALCULATIONS AND ANALYSIS OF MEASURED AND CALCULATED DATA 8.1 INTRODUCTION TO THE DATA ANALYSIS METHODOLOGY The peculiarities of this project, distinguishing it from similar benchmark intercomparisons, are the comparison of not only integral values but also of spectra and the inclusion of gamma spectra as well as of thermal neutrons connected with a realistic modeling of the WWER RPV environment. Advantages of LR-0 experiment relative to existing RPV benchmarks are: Large penetrations are modeled in a realistic model (i.e. in the VENUS benchmarks the maximal radial material thickness outside the core would be equivalent to WWER-1000 baffle only, not the RPV). Neutron and gamma spectrum measurements - no equivalent presently available. Specific benchmark for WWER, where the outer WWER-1000 and WWER-440 regions are modeled in a realistic way. As results of Work-Packages 1 and 2 of the REDOS project, available and new experimental data were specified together with all reactor data needed for the calculation analysis of the experiments. In Work-Package 3 the experimental benchmark results on neutron and gamma spectra measurements in the Mock-ups of the WWER-1000 and WWER-440 reactors at the LR-0 were compared with different independent calculations. The found discrepancies among the different calculations as well as between calculations and experiments were analyzed. The three following Mock-ups with the core simulating the steady state power distribution (after several fuel cycles) have been considered: WWER-1000 Mock-up: WWER-1000 with a simple core loading. WWER-440 Mock-up No 1: WWER-440 with standard core loading,, and with maximum leakage neutron flux density along the Mock-up symmetry axis. WWER-440 Mock-up No 2: WWER-440 with dummy steel assemblies at the core boundary, with minimum leakage neutron flux density along the Mock-up symmetry axis. Seven institutions from 5 countries (Bulgaria, Czech Republic, Germany, Hungary and Spain) delivered calculation results for the WWER-1000 Mock-up and six institutions from 4 countries performed calculations for one or both WWER-440 Mock-ups. The following abbreviations have been used for the participating institutions delivering calculation results: 24

26 FANP UJV SKODA INRNE SEA FZR AEKI. Framatome ANP, Erlangen. Nuclear Research Institute, Řež, Czech Republic. Škoda, Plzen, Czech Republic. INRNE, Sofia, Bulgaria. SEA, Madrid, Spain. Forschungszentrum Rossendorf, Dresden. KFKI AEKI Budapest, Hungary. More than eight different computational schemes were used, covering the most important methods used for pressure vessel dosimetry and shielding. The codes used were the stochastic codes MCNP and TRAMO and the deterministic codes ANISN/DORT and TORT: MCNP DORT Synthesis The most widely used continuous energy Monte Carlo code [22]. It had been used in a fixed source mode by 3 participants and in a criticality mode by the UJV. SEA used its standard procedure of calculating assembly-to-assembly contributions to the detector signals. Different strategies to reduce the statistical errors have been applied. A widely used method for routine calculations - a 3-D synthesis of (R- Θ)-, (R-Z)- and R-calculations with the discrete ordinates codes DORT and ANISN [23]. TORT A 3-D discrete ordinates code [23]. TRAMO A group data Monte Carlo code developed in the FZR [24] They were associated with different data libraries and data preparation schemes. The BUGLE 47n/20g group structure was used for comparisons of calculated spectra as well as for comparison with experimental spectra. Some participants compared calculation results with experiments also in a finer group structure. The following data libraries have been used: ENDF/B-VI BUGLE-96 The 6th and last version of the evaluated nuclear data library ENDF/B in different revision states. A group data library with 47 neutron and 20 gamma groups based on ENDF/B-VI rev. 3 not considering neutron up-scattering in the thermal region. BUGLE-96T BUGLE-96 with thermal up-scattering [25]. 25

27 BGL1000 A problem-oriented group library for WWER-1000 type reactors [31] with BUGLE group structure. BGL440 A problem-oriented group library for WWER-440 type reactors [31] with BUGLE group structure. TRAMO- 47/20 TRAMO- 640/94 A group data library with reactors with BUGLE group structure generated by NJOY based on ENDF/B-VI rev. 7 and 8. A group data library with 640 neutron and 94 gamma groups generated by NJOY based on ENDF/B-VI rev. 7 and 8. All calculations and comparisons were done for the measurement points at different radii in the central plane along the azimuthal symmetry axes of the Mock-ups. The measurements of the azimuthal and axial fluence distributions were not considered, as preliminary investigations had shown that the changes expected and actually found were in the range of the measurement errors. In the following the results of the analysis are given for the 3 Mock-ups separately. In Tables and Figures the calculation variants are marked by the following abbreviations: FANP S UJV M SKODA T INRNE M INRNE T SEA M FZR 47 FZR 640 AEKI M For DORT Synthesis / BUGLE-96, FRAMATOME ANP, Erlangen. For MCNP / ENDF/B-VI, Nuclear Research Institute, Řež, Czech Republic. For TORT - BUGLE-96T Škoda, Plzen, Czech Republic. For MCNP - ENDF/B-VI, INRNE, Sofia, Bulgaria. For TORT - BGL1000, INRNE, Sofia, Bulgaria. For MCNP - ENDF/B-VI, SEA Madrid, Spain. For TRAMO / 47n/20g ENDF/B-VI, FZ Rossendorf, Dresden, Germany. For TRAMO / 640n/94g ENDF/B-VI, FZ Rossendorf, Dresden, Germany. For MCNP - ENDF/B-VI, KFKI AEKI Budapest, Hungary. 8.2 ANALYSIS OF CALCULATION RESULTS FOR THE WWER-1000 MOCK-UP Points of measurements and calculations The calculations and analyses were performed for the measurement points in the central plane of the Mock-up along its symmetry axis in the radial positions according to Table 6 [7]. 26

28 Table 6. Location of measuring points for neutron spectrum Point No Position Displacer position 0 Central dry channel in the assembly No At the barrel 2 At the barrel in the displacer At the barrel 21 In the displacer Displacer 5 cm from the barrel 22 In the displacer Displacer 10 cm from the barrel 23 In the displacer Displacer 15 cm from the barrel 3 Before RPV At the barrel 4 5 cm thickness of RPV At the barrel 5 10 cm thickness of RPV At the barrel 6 15 cm thickness of RPV At the barrel 7 Behind RPV At the barrel As it can be seen in Table 7 all participants performed calculations of fluxes and DPA rates for a total core neutron source of 1 neutron/s. That allowed to prove the ability of the participants to calculate flux spectra at an absolute scale. Participant Table 7: List of calculation results for the WWER-1000 Mock-up Code/Library Detector Flux integrals and DPA 47-group n-spectra 20-group γ-spectra FANP DORT Synthesis / BUGLE-96 no all all all FANP DORT Synthesis / BUGLE-96 yes p2, p3 p2, p3 p2, p3 UJV MCNP / ENDF/B-VI no all all all SKODA TORT / BUGLE-96T no all all all SKODA TORT / BUGLE-96T yes all all all INRNE TORT - BGL1000 no INRNE MCNP - ENDF/B-VI no p.2,3,4,5, 6,7 p.2,3,4,5, 6,7 p.2,3,4,5, 6,7 p.2,3,4,5, 6,7 p.2,3,4,5, 6,7 p.2,3,4,5, 6,7 INRNE MCNP - ENDF/B-VI yes n: p.2,3,4 SEA MCNP - ENDF/B-VI no all all all AEKI MCNP - ENDF/B-VI no all all,>0.111 all,>0.2 MeV MeV FZR TRAMO / 47n/20g ENDF/B-VI no all all all FZR TRAMO / 640n/94g ENDF/B-VI no all all all

29 8.2.2 Comparison of the calculation results In Table 8 [13] the relative standard deviations (RSD, mean square root deviation) of the participants results, which characterize to some degree the accuracy of the results, are gathered. As we can see, the fast neutron region with RSD between 4 and 12%, an excellent to good agreement between participants results could be stated, with the exception of point 0 with RSD up to 16%. Possibly, the applied calculation procedures were better adjusted to the RPV region than to the core. Good agreement existed at all positions for the compared neutron spectral indices. Considerably higher discrepancies between participants results were found for the thermal flux (RSD 17%-49%) and the gamma flux integrals and DPA (RSD 10%-35%). Thermal flux discrepancies would reduced by 5-30% if the FANP results, obtained with BUGLE-96 data without consideration of up-scattering would be excluded from the comparison. As an example, Figures 8-11 [13] show the neutron and gamma flux spectra per one source neutron/s as calculated by the participants for point 2 (surveillance position) and point 7 (behind the RPV). The spectra were given in differential form as flux per lethargy unit in dependence on the particle energy in a logarithmic scale. The abscissa points were the midpoints of the lethargy group intervals. This display format ensured that the area under the spectrum curve in a particular energy region was proportional to the number of particles/cm 2 s in that energy region. Table 8: Relative standard deviation of participants absolute integral results from their mean values E thr, [MeV] Relative standard deviation of participants results from their mean values DPA P0 P2 P21 P22 P23 P3 P4 P5 P6 P7 En<0.414 ev En> En> En> En> En> En> DPAn Eg> Eg> Eg> Eg> DPAg >.4979/>3.012MeV-n >1.003/>3.012MeV-n