Belgian Nuclear Research Centre SCK Research Articles

Performance assessment and further optimization of the activation based criticality dosimetry system at the Belgian nuclear research centre SCK CEN

In places where fissile materials are being handled, the risk of a criticality accident can never be absolutely eliminated. Criticality accidents require specialized dosimetry techniques, which differ markedly from those used in routine radiological protection. This is caused by the fact that one should be able to measure very high doses up to 10 Gy in mixed gamma and neutron fields with very high instantaneous dose rates up to 105 Gy/s and complex energy spectra. The criticality dosimetry system at the Belgian Nuclear Research Centre SCK CEN is accredited according to ISO 17025 by the Belgian accreditation body BELAC and is used to monitor about 4000 nuclear workers. The system is based on a dedicated criticality dosemeter with four activation detectors (gold foils with and without cadmium shielding, an indium foil and sulphur pellets) and an iterative unfolding algorithm starting from a guess neutron fluence energy spectrum. This work describes the design of the SCK CEN criticality dosemeters and the iterative unfolding algorithm. Furthermore, it also discusses two new developments. Firstly, effective neutron reaction cross sections were calculated by MCNP6.2 Monte Carlo radiation transport simulations to be used in the unfolding algorithm. These effective cross sections take into account the effects of the dosemeter, the potential presence of the body and the neutron angle of incidence. These effects were demonstrated to be important, especially for the gold foils and if the dosemeter is worn against the body. Secondly, a Bayesian unfolding algorithm based on a parametric description of the neutron fluence energy spectrum was developed and tested. The results of this new algorithm were in good agreement with the iterative algorithm. The uncertainty assessment was more realistic with the Bayesian algorithm because it includes also the uncertainty associated with the unfolding process. However, for spectra that cannot be well described by the parametric function, the Bayesian algorithm can lead to significant under- or overestimations. In case sufficient information is available, it is better to make use of the iterative algorithm with a realistic guess spectrum to start from. Finally, this work also demonstrated the compliance of the SCK CEN criticality dosimetry system with the IAEA criteria for criticality dosimetry systems.

Validating Geant4 with Monte Carlo simulations in the context of nuclear disarmament verification

In this paper neutron transport simulation results of experimental configurations used during the 2019 IPNDV measurement campaign at the Belgian Nuclear Research Centre SCK CEN in Mol are presented. As these are a good basis for benchmarking simulations, results of the Monte-Carlo simulation codes Geant4, SCALE/KENO-VI, MCNP and openMC are compared with the aim of validating Geant4. With fissile material present as plutonium–uranium mixed oxide, (α,n) emissions are included in the simulations. A Geant4 extension to calculate neutron multiplication factors keff was developed and applied to the IPNDV configurations and a set of OECD/NEA benchmark experiments. Neutron fluxes of the IPNDV configurations through a reference volume are simulated. The Geant4 (α,n) simulation toolkit SaG4n is compared to the SCALE code ORIGEN and small deviations are observed. Geant4 calculations of keff for criticality benchmark experiments deviate in the mean by ＋0.4%, for MCNP and openMC deviations are ≤6%, for KENO-VI ≤8%. Geant4’s total neutron fluxes through a reference volume next to the IPNDV configurations agree within a margin of ±5% with MCNP and openMC and −14 % with KENO-VI. Considerable differences between all codes were observed for thermal neutron scattering but Geant4 results are still within the other codes’ variability. Differences to Geant4 for epithermal energies can be addressed to a not yet included “Doppler Broadening Rejection Correction”. By comparing Geant4 simulations of a complex set-up with various reference codes and demonstrating its good performance as well as some discrepancies, this study contributes to the validation of Geant4 neutron physics in fissile material systems and for nuclear disarmament verification simulations.

Study of Reactor Core Loading Monitoring at the GUINEVERE Facility

The potential use of a pulsed neutron source (PNS) to measure reactivity during nuclear fuel loading as a means to prevent core loading errors has been studied at the GUINEVERE facility. This facility couples the deuteron accelerator GENEPI-3C to the fast neutron subcritical reactor VENUS-F at the Belgian Nuclear Research Center SCK·CEN. The 14-MeV neutrons are produced in the reactor core center via fusion reactions. PNS experiments were performed in five reactor configurations corresponding to the different loading steps of VENUS-F. The evolution of the neutron flux during these PNS experiments was measured by several 235U fission chambers in various positions in the inner and outer reflector and analyzed using the area-ratio method. The results show that, despite strong spatial effects, a strong correlation between the reactivity values given by the area-ratio method and some reference reactivity values remains throughout the reactor unloading. Monte Carlo simulations were first validated by comparison with the data and then used to investigate the sensitivity of the method to a core loading error. First results show that some loading errors could be experimentally detected using a PNS.

Hunting for the correct pressure drop in a scaled reactor pool: Effect of geometry, mesh resolution, turbulence model and mass flow

The E-SCAPE (European SCAled Pool Experiment) facility at the Belgian nuclear research center SCK•CEN is a thermal hydraulic 1/6th-scale model of the MYRRHA reactor with an electrical core simulator and cooled by LBE. The E-SCAPE facility provides experimental feedback to the designers on the forced and natural flow patterns, flow mixing and stratification during operational and accidental conditions. In addition, it provides valuable data for validation of the computational techniques for their use with LBE and in the design of MYRRHA. In terms of momentum field, one of the most important integral quantity to reproduce is the pressure drop in the system which has a direct consequence on the evolution of the mass flows and determines the free-surface levels in transient scenarios. In this paper, a detailed CFD study is performed on the flow in the E-SCAPE core region with the aim on predicting the correct pressure drop. Different numerical representation of the core geometry have been modeled and subjected to validation against experimental results from E-SCAPE. Special attention has been addressed to the choice of geometric simplifications and numerical models for accurate representation of the core of the E-SCAPE pool facility.

Validation of CFD analyses against pool experiments ESCAPE

MYRRHA is a pool-type research reactor under design at the Belgian nuclear research center SCK·CEN. The multi-purpose fast-spectrum MYRRHA reactor is cooled with lead–bismuth eutectic (LBE). ESCAPE is a thermal hydraulic 1/6th-scale model of the MYRRHA reactor developed within the European framework programs THINS and MYRTE.The experiments from ESCAPE are crucial for the design and licensing of MYRRHA. First of all, these experiments can provide information on the flow patterns and heat transport in the different compartments of the reactor. Secondly, the data from the ESCAPE experiments can be used for qualification of thermal hydraulic computer codes that will be applied for analyzing the flow and heat transport in the MYRRHA reactor. Considering the rare and typical conditions in heavy liquid cooled nuclear reactors, thorough validation of these codes against relevant experimental data is a prerequisite before application.This paper presents the approach followed by NRG to model the flow and heat transport in ESCAPE using the Computational Fluid Dynamics (CFD) code STAR-CCM+. CFD analyses for different forced flow conditions are presented and compared with experimental results. The predicted temperature distribution in the pool, core and plena agrees well with the temperature measurements from ESCAPE.

Fusion target station on MYRRHA facility: Baseline concept

The Belgian Nuclear Research Centre SCK CEN launched the 1st phase implementation of the MYRRHA reactor. This will deploy the construction of 100 MeV proton accelerator and complimentary research facilities including a dedicated station for irradiation of fusion materials (Fusion Target Station - FTS). This paper presents a baseline pre-conceptual design and analysis of the irradiation conditions for the most common in-vessel fusion materials i.e. tungsten and Eurofer97 steel. FTS will allow a direct irradiation by protons and/or irradiation via spallation neutron source. The former concept requires application of small specimen test techniques (SSTT), while the latter enable large irradiation volume with a limit damage rates.

CFD analyses of the European scaled pool experiment E-SCAPE

The European Scaled Pool Experiment (E-SCAPE) is a 1/6th-scale thermal hydraulic model of the MYRRHA reactor and has been developed within the European framework programs THINS and MYRTE. MYRRHA is a multi-purpose hybrid pool-type research reactor cooled with lead-bismuth eutectic (LBE) that is under design at the Belgian nuclear research center SCK•CEN.The experiments from the E-SCAPE facility are essential for the design and licensing of MYRRHA. On the one hand, the experimental data from E-SCAPE directly provide information to the designers on the flow patterns and heat transport in the LBE in the upper and lower plena of the reactor. On the other hand, the experimental data from E-SCAPE enable qualification of thermal hydraulic computer codes used for analyzing the flow and heat transport in the MYRRHA reactor. The accuracy of codes rely on the underlying modelling approaches, whose validity can only be assessed by comparing them with relevant experimental data. Given the rare conditions present in heavy liquid cooled nuclear reactors, the possibility to compare complete pool simulations with experimental data is a unique opportunity.This paper presents the modelling approaches and first results of the pre-test analyses performed for E-SCAPE by NRG, SCK•CEN and VKI using Computational Fluid Dynamics (CFD). The CFD analyses of NRG are performed with the STAR-CCM + code, at SCK•CEN with the ANSYS CFX code and at VKI OpenFOAM is used. A first comparison of the results obtained with three different CFD codes for flow and heat transport in the E-SCAPE pool is presented and, where possible, validated against experimental results.

Coupled system thermal-hydraulic/CFD analysis of a protected loss of flow transient in the MYRRHA reactor

Within the MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) project, the Belgian Nuclear Research Centre SCK•CEN is developing and designing a flexible irradiation facility, configured as an accelerator driven system (ADS) with a MOX fueled reactor core, able to operate in both critical and sub-critical modes. The system design features a compact pool-type primary cooling system operating with molten Lead-Bismuth Eutectic (LBE). With regard to the thermal-hydraulic design and safety assessment of the installation, a major challenge is represented by the complex coolant flow field characterizing the large open regions of the primary vessel, namely the cold and hot plenum, with the presence of pronounced three-dimensional phenomena that may impact the evolution of accidental transients such as loss of flow events. In order to have a realistic representation of such effects, a coupled system thermal-hydraulic/CFD model of the MYRRHA reactor is developed and presented in this paper. The proposed multi-scale methodology, which couples the 1D system code RELAP5-3D to the CFD code FLUENT, is based on domain decomposition and a novel implicit numerical scheme is developed. The coupled reactor-scale computational model is applied in this work to the analysis of a postulated protected loss of flow (PLOF) accident, and preliminary validated against RELAP5-3D stand-alone solution data. The results of the analysis were found in agreement, demonstrating the capability of the tool to perform integral simulation taking into account 3D flows and local phenomena.

Extension and application on a pool-type test facility of a system thermal-hydraulic/CFD coupling method for transient flow analyses

The development of multi-scale modeling capabilities through the use of coupled system thermal-hydraulic and CFD codes is addressed in this paper, which presents the extension and application of a partitioned, domain decomposition computational method coupling the 1D code RELAP5-3D and the CFD code FLUENT. An implicit coupling scheme for the solution of the flow field, based on a Quasi-Newton algorithm, is developed and tested on a case with multiple coupling interfaces. Moreover, modeling capabilities of the tool are enhanced through the implementation of thermal coupling to compute conjugate heat transfer phenomena. In this work, a coupled model has been developed for the analysis of the pool-type test facility E-SCAPE, currently under commissioning at the Belgian Nuclear Research Centre SCK•CEN. The installation represents a thermal-hydraulic scaled model of the MYRRHA reactor, with an electrical core simulator, cooled by Lead-Bismuth Eutectic (LBE). The coupling tool is applied on the analysis of a total loss of flow (LOF) transient, involving the complex transition from forced to natural circulation flow in the primary system. The simulation results, compared against stand-alone system thermal-hydraulics (STH) simulation data, highlighted that the transient behavior of a pool-type system in natural circulation is characterized by complex multi-dimensional flow and temperature fields, difficult to predict by 1D STH codes. The study also confirmed the capability of the developed tool to predict the impact of such phenomena on the system behavior, and to capture the development of thermal stratification in a plenum at low flow condition. The planned experimental tests will be used for the validation of the tool, in the view of its use for design and licensing activities.

Nuclear data uncertainty propagation on a sodium fast reactor

Uncertainties on criticality parameters and reactivity effects in a SFR (Sodium Fast Reactor) were calculated and characterized. An SFR model loaded with ≈10% MAs (Minor Actinides) was chosen as test case of a burner reactor, its core geometry was implemented in MCNP-5 according to benchmarking specifications. Monte Carlo random sampling on the ENDF/B-VII.1 neutron-induced nuclear data was performed using the SANDY code developed at the Belgian nuclear research center SCK•CEN. SANDY was proved to be a successful tool for uncertainty quantification in the model case of a MA burner SFR. The results show that despite the high content in MAs, the contribution of U-238 data is still dominant in most instances, especially the inelastic scattering cross section data. Minor actinides together with Pu make up to 33% of the total calculated keff uncertainty, which amounts to ±1345pcm of reactivity. Relative uncertainty greater than ±3.47% was estimated for βeff with significant contribution from the prompt multiplicity data of U-238 and Pu-239. Relative uncertainty on the Doppler reactivity worth was found to be at least ±12.6%. Coolant voiding in different core zones yields reactivity worth with relative uncertainties whose lower bounds range from ±6.37% to ±22.47% depending on the voiding conditions, with important contributions due to the Na-23 nuclear data uncertainty.

Estimate of the reactivity of the VENUS-F subcritical configuration using a Monte Carlo MSM method

The GUINEVERE project was launched in 2006 in order to study the feasibility of transmutation in Accelerator Driven subcritical Systems (ADS). This facility, hosted at the Belgian Nuclear Research Center SCK·CEN, couples the fast subcritical lead reactor VENUS-F with an external neutron source provided by the GENEPI-3C accelerator.In order to test on-line subcriticality monitoring techniques, an independent measurement of the reactivity of VENUS-F was achieved using the Modified Source Multiplication Method (MSM). The unknown reactivity of the subcritical configuration of interest was determined by comparing detector count rates driven by an external neutron source in this configuration with those obtained in a slightly subcritical configuration whose reactivity was determined by rod drop experiments. To account for spatial effects, MSM correction factors were calculated for all the detectors using the Monte Carlo neutron transport code MCNP. The corrected reactivity values of all the detectors were found to be consistent. This led to a final estimate of −5.28±0.13$ for the reactivity of VENUS-F.The MSM factors calculated with MCNP were found to be insensitive to the assumptions made for modeling the reactor and the detectors. Thus the MSM method appears to be a robust technique for measuring large subcriticality values with good accuracy.

VALIDATION OF A MONTE CARLO MODEL OF THE FORK DETECTOR WITH A CALIBRATED NEUTRON SOURCE

The investigation of experimental methods for safeguarding spent fuel elements is one of the research areas at the Belgian Nuclear Research Centre SCK•CEN. A version of the so-called Fork Detector has been designed at SCK•CEN and is in use at the Belgian Nuclear Power Plant of Doel for burnup determination purposes. The Fork Detector relies on passive neutron and gamma measurements for the assessment of the burnup and safeguards verification activities. In order to better evaluate and understand the method and in view to extend its capabilities, an effort to model the Fork detector was made with the code MCNPX. A validation of the model was done in the past using spent fuel measurement data. This paper reports about the measurements carried out at the Laboratory for Nuclear Calibrations (LNK) of SCK•CEN with a 252Cf source calibrated according to ISO 8529 standards. The experimental data are presented and compared with simulations. In the simulations, not only was the detector modeled but also the measurement room was taken into account based on the available design information. The results of this comparison exercise are also presented in this paper.

Review of the natC(n, γ) cross section and criticality calculations of the graphite moderated reactor BR1

A review of the experimental data for natC(n,γ) and 12C(n,γ) was made to identify the origin of the natC capture cross sections included in evaluated data libraries and to clarify differences observed in neutronic calculations for graphite moderated reactors using different libraries. The performance of the JEFF-3.1.2 and ENDF/B-VII.1 libraries was verified by comparing results of criticality calculations with experimental results obtained for the BR1 reactor. This reactor is an air-cooled reactor with graphite as moderator and is located at the Belgian Nuclear Research Centre SCK•CEN in Mol (Belgium). The results of this study confirm conclusions drawn from neutronic calculations of the High Temperature Engineering Test Reactor (HTTR) in Japan. Furthermore, both BR1 and HTTR calculations support the capture cross section of 12C at thermal energy which is recommended by Firestone and Révay. Additional criticality calculations were carried out in order to illustrate that the natC thermal capture cross section is important for systems with a large amount of graphite. The present study shows that only the evaluation carried out for JENDL-4.0 reflects the current status of the experimental data.

The Use of μCT and ESEM in the Study of the Osmosis-Induced Water Uptake by Eurobitum Bituminized Radioactive Waste

Laboratory water uptake tests are performed at the Belgian Nuclear Research Centre SCK•CEN to obtain insight into the hydromechanical behavior of Eurobitum bituminized radioactive waste under geological disposal conditions. Small nonradioactive and radioactive Eurobitum samples are hydrated in restricted swelling conditions (i.e., nearly constant volume conditions and constant stress conditions). Microfocus X-ray computer tomography (μCT) proves to be a very suitable technique to follow up the ingress of water in the samples. μCT analyses demonstrate that, under the studied hydration conditions, the water uptake by Eurobitum samples is a diffusion controlled process. A characterization of the partially leached samples with environmental scanning electron microscopy (ESEM) shows that the hydration of salt crystals and the subsequent dilution of the salt solution result in an increase in pore size that is limited to a few tens of μm in restricted swelling conditions. The μCT and ESEM analyses allow improvement in the understanding of water uptake by Eurobitum in restricted swelling conditions. In this article we discuss the μCT and ESEM analyses of nonradioactive Eurobitum samples that were hydrated for 2 to 4 years at a constant stress of 1, 22, 33, and 44 bar or in nearly constant volume conditions.

Osmosis-induced water uptake by Eurobitum bituminized radioactive waste and pressure development in constant volume conditions

The chemo-hydro-mechanical (CHM) interaction between swelling Eurobitum radioactive bituminized waste (BW) and Boom Clay is investigated to assess the feasibility of geological disposal for the long-term management of this waste. These so-called compatibility studies include laboratory water uptake tests at the Belgian Nuclear Research Center SCK•CEN, and the development of a coupled CHM formulation for Eurobitum by the International Center for Numerical Methods and Engineering (CIMNE, Polytechnical University of Cataluña, Spain).In the water uptake tests, the osmosis-induced swelling, pressure increase and NaNO3 leaching of small cylindrical BW samples (diameter 38mm, height 10mm) is studied under constant total stress conditions and nearly constant volume conditions; the actual geological disposal conditions should be intermediate between these extremes. Two nearly constant volume tests were stopped after 1036 and 1555days to characterize the morphology of the hydrated BW samples and to visualize the hydrated part with microfocus X-ray Computer Tomography (μCT) and Environmental Scanning Electron Microscopy (ESEM). In parallel, a coupled CHM formulation is developed that describes chemically and hydraulically coupled flow processes in porous materials with salt crystals, and that incorporates a porosity dependent membrane efficiency, permeability and diffusivity.When Eurobitum BW is hydrated in (nearly) constant volume conditions, the osmosis-induced water uptake results in an increasing pressure to values that can be (in theory) as high as 42.8MPa, being the osmotic pressure of a saturated NaNO3 solution. After about four years of hydration in nearly constant volume water uptake tests, pressures up to 20MPa are measured. During this hydration period only the outer layers with a thickness of 1–2mm were hydrated (as derived from μCT and ESEM analyses), and only about 10–20% of the initial NaNO3 content was released by the samples. In the studied test conditions, the rates of water uptake and NaNO3 leaching are low because of the low porosity, and thus low permeability, of the hydrated BW samples in combination with a highly efficient semi-permeable bitumen membrane. In contrast to the hydration in free swelling conditions, the increase in porosity is limited by the high pressures in the nearly constant volume tests. Furthermore, at the interface with the stainless steel filters, a low permeable re-compressed bitumen layer is formed, as observed on the ESEM images.The experimental results of pressure increase and NaNO3 leaching, as well as observations on μCT and ESEM images (e.g. compression of leached layers, high dissolved NaNO3 concentration in hydrated BW after about four years), were reproduced rather successfully by the coupled CHM formulation for Eurobitum BW. A long-term model prediction of the evolution of the osmosis-induced pressure in the nearly constant volume tests shows that the pressure would reach a maximal value of about 20MPa after about 5.5years, after which the pressure would start to decrease. After 10,000days (∼27years) the pressure would have decreased to a value of ∼2MPa.

The osmosis-induced swelling and NaNO3 leaching of radioactive and artificially aged Eurobitum bituminized waste

ABSTRACTThe behavior of Eurobitum bituminized radioactive waste under geological disposal conditions is studied in water uptake tests at the Belgian Nuclear Research Centre SCK•CEN to assess the feasibility of geological disposal for the long-term management of this waste. The hydro-mechanical behavior of this waste is affected by the continuous evolution of the rheological properties of bitumen due to radio-oxidation (i.e. ageing).The effect of the bitumen ageing degree on the kinetics of the water uptake, swelling and NaNO3 leaching is investigated in water uptake tests with ∼30 years old radioactive samples and inactive samples that were artificially aged. The first results of swelling and NaNO3 leaching of (i) radioactive samples that have been hydrated for more than 2 years at a constant total stress of 2.2 MPa, and (ii) a thermally aged sample that has been hydrated for ∼1.5 years under nearly zero effective stress conditions, revealed lower swelling and higher leach rates for these samples compared to non-aged samples. The effect of ageing on the osmotic efficiency of bitumen as a semi-permeable membrane is less pronounced when swelling of the samples is limited, and changes in time, probably because of the formation of low porosity layers, which seems to mask the difference in bitumen membrane efficiency of aged and non-aged bituminized waste.

Design of a planar probe diagnostic system for plasmatron VISIONI and its application for the study of deuterium retention in W–Ta alloys

One of the issues still to be clarified with respect to the possible use of tungsten and tungsten alloys as first wall material in fusion reactors is the retention of hydrogen isotopes under fusion-relevant conditions, such as the presence of helium ash, neutron irradiation, thermal shock. The low flux, high temperature plasma simulator VISIONI, being currently constructed at the Belgian nuclear research center SCK·CEN, will have the unique ability to handle radioactive materials – neutron irradiated samples and tritium plasma. Currently it is being tested with non-irradiated samples and deuterium plasma. To monitor the plasma parameters near the investigated specimen, a planar probe system has been developed. It was demonstrated that the plasmatron is able to provide conditions relevant for the ITER first wall. Investigations of deuterium retention in W–Ta alloys have been performed, using thermodesorption spectroscopy. A hypothesis of the existence of three kinds of trapping sites for deuterium in W–Ta alloys, based on deuterium release studies, is suggested.

The methodology followed in Belgium to investigate the compatibility with geological disposal of Eurobitum bituminized intermediate-level radioactive waste

AbstractIn Belgium, Eurobitum intermediate-level long-lived bituminized radioactive waste containing large amounts of NaNO3, which is a hygroscopic and soluble salt, is to be disposed of in an underground repository in a geologically stable clay formation. The Boom Clay is studied as a potential host formation because of its favourable properties to limit and delay the migration of the leached radionuclides and other contaminants (heavy metals, NaNO3, organic molecules) to the biosphere. The emplacement of the bituminized waste will induce multiple processes that could have a significant effect on the key properties of the clay. Because several of these processes are interdependent, the study of the compatibility of Eurobitum with geological disposal is complex. To structure the research and to identify possible knowledge gaps, the Belgian Radioactive Waste Management Agency ONDRAF/NIRAS developed a new methodology based on safety functions and safety statements. In this paper, this methodology is briefly explained, with reference to the disposal of Eurobitum. Experimental results obtained at the Belgian Nuclear Research Centre SCK•CEN are presented and discussed in the light of the safety functions and safety statements approach. The importance of the interdependence of the processes is highlighted. Special attention is given to the evolution of the disposal design as a result of the improved understanding of key processes.

Assessment of americium and curium transmutation in magnesia based targets in different spectral zones of an experimental accelerator driven system

The potential to incinerate minor actinides (MA) in a sub-critical accelerator-driven system (ADS) is a subject of study in several countries where nuclear power plants are present. The performance of the MYRRHA experimental ADS, as to the transmutation of Am and Cm in the inert matrix fuel (IMF) samples consisting of 40vol.% (Cm0.1Am0.5Pu0.4)O1.88 fuel and 60vol.% MgO matrix with a density of 6.077gcm−3 in three various spectrum regions, were analysed at the belgian nuclear research centre SCK·CEN. The irradiation period of 810 effective full power days (EFPD) followed by a storage period of 2 years was considered. The ALEPH code system currently under development at SCK·CEN was used to carry out this study. The total amount of MA is shown to decrease in all three considered cases. For Am, the decrease is the largest in the reflector (89% decrease) but at the cost of a net Cm production (92% increase). In the two other positions (inside the core region), 20–30% of Am has disappeared but with a lower production of Cm (between 7% and 11%). In the reflector, a significant build-up of long-lived 245Cm, 246Cm, 247Cm and 248Cm was also observed while the production of these isotopes is 10–1000 times smaller in the core. The reduction of the Pu content is also the highest in the reflector position (41%). In the other positions the incinerated amount of Pu is much smaller: 1–5%.