High Flux Reactor In Petten Research Articles

Microstructural evolution investigation of irradiated oxide fuels by means of electron backscattered diffraction

In this paper, both irradiated UO2, together with a fresh UO2.001 sample from the same batch as a reference, and irradiated FR-MOX U0.55Pu0.45O1.99 (TRABANT 2/3) fuel were studied using Electron Back Scattered Diffraction (EBSD). Irradiation of both oxide fuels was carried out at the High Flux Reactor (HFR) Petten, the UO2 at a moderate maximum Linear Heat Rate (LHR) of 200 W/cm to a burn-up of 5.4 % FIMA (fission per initial metal atom) or 50 MWd/kgUO2, and the MOX at a high maximum Linear Heat Rate (LHR) of 374 W/cm to a burn-up of 1.7% FIMA. In the irradiated UO2, the grain growth was observed in the fuel centre. The irradiated MOX (U, Pu)O2 showed the typical FBR fuel microstructure zones, i.e. as-fabricated (at the fuel edge), equi-axed and columnar grains as observed from the fuel-clad region towards fuel central hole.The LAGB (Low Angle Grain Boundary), 3° < θ < 15°, as an initial state for the HBS structure was observed on both UO2 and MOX. This restructuring mechanism for both oxide fuels is proposed to follow the well-known polygonization mechanism in which an arrangement of dislocations at the grain boundaries occurs.

Radiolytic Production of Fluorine Gas from MSR Relevant Fluoride Salts

Radiolytic fluorine gas production at temperatures of 40°C to 60°C was investigated for the fluoride salts LiF, BeF2, UF4, ThF4, and 71.7LiF-16BeF2-12.3UF4 (FliBe-UF4) by gamma irradiation of powdered samples using spent fuel elements from the High Flux Reactor (HFR) Petten as the irradiation source; work of a similar nature was previously performed at Oak Ridge National Laboratory in the period 1965 to 1995. Gamma irradiation was conducted for just over 41 days, with total absorbed gamma dose ranging from ~45 MGy for the lightest salts to ~170 MGy for ThF4 and UF4. By measuring the gas pressure within salt-filled capsules during irradiation, it was possible to quantify radiolytic gas production for all salt samples except UF4. Production rates are reported as the salt G-values, measured as number of fluorine molecules produced per 100 eV of energy absorbed (molecules F2/100 eV). The G-values of the salts were found to be G(LiF) ~0.004, G(BeF2) ~0.009, G(ThF4) ~0.021, and G(FLiBe-UF4) ~0.005.

Post-irradiation microstructural examination of EUROFER-ODS steel irradiated at 300°C and 400°C

The effect of neutron irradiation on the microstructure of EUROFER-ODS alloy reinforced with an addition of 0.5% Y 2 O 3 was analyzed by transmission electron microscopy . The irradiation of up to 16.2 dpa damage dose at 300°C and 400°C was performed in Petten High Flux Reactor. This study focuses on the imaging, identification, and quantitative analysis of interstitial loops and voids located on ODS particles. The nano-sized loops and point defects with b½〈111〉 Burgers vectors are preferentially located near structural defects such as the line dislocations or impurities in the material irradiated at 300°C. Homogeneously distributed loops having Burgers vectors of b½〈111〉 and b〈100〉 types were found in the material irradiated at 400°C. The obtained quantitative data on radiation-induced defects such as size, number density and relaxation volume were compared with EUROFER97 irradiated at the same conditions. It was found that the occurrence of radiation induced defects depends on the local number density and the size distribution of the ODS particles and not only on the irradiation damage and temperature.

Influence of Ni-Mn contents on the embrittlement of PWR RPV model steels irradiated to high fluences relevant for LTO beyond 60 years

Understanding the irradiation induced degradation of the reactor pressure vessel (RPV) steels at high fluences becomes more important in view of safe long term operation (LTO) of existing nuclear power plants. Specifically, the role of Ni, Mn and Si and associated synergetic effects on embrittlement behaviour need to be further investigated to understand accelerated embrittlement at high neutron fluences (ca. 1 × 10 20 n.cm −2 , E>1 MeV) observed in certain types of RPV steels. In addition, it is required to verify the validity of existing embrittlement trend curves (ETCs) at high fluence values. To address these issues, NRG and JRC performed a unique irradiation experiment called LYRA-10 in the High-Flux Reactor (HFR) Petten. In LYRA-10 a variety of specimens (e.g. tensile, Charpy, mini-Charpy and microscopy) of RPV steels (base metal and welds) with systematic variations in Ni, Mn and Si contents, while keeping the base composition representative of VVER-1000 and PWR RPV steels, have been irradiated for 16 HFR cycles to achieve the nominal fast neutron fluence of 1.11 × 10 20 n.cm −2 at a nominal temperature of 286 °C. A large post-irradiation examination campaign on LYRA-10 specimens has been initiated within the STRUMAT-LTO, which is an international collaboration project co-funded by EURATOM. The research reported in this paper includes results from tensile and mini-Charpy impact testing of four PWR RPV model steel specimens irradiated in LYRA-10, which is performed as part of STRUMAT-LTO in-kind pilot project before the start of the EURATOM co-funded project. The combined influence of Ni and Mn on embrittlement behaviour, characterized by irradiation hardening and shifts in transition temperature (T k ) properties, has been evaluated. The measured shifts in T k values for all four model steels have been compared with predicted values from two selected ETCs in order to check their validity for these steels at the above high neutron fluence values.

INTRODUCTION OF OSCAR-4 AT THE HIGH FLUX REACTOR (PETTEN)

Since 2005 the nodal diffusion based code system, OSCAR-3, was used for reactor support calculations of operational cycles of the High Flux Reactor in Petten, The Netherlands. OSCAR uses a two-step deterministic calculation, in which homogenized cross sections are generated in lattice environments using neutron transport simulations, and then passed to a nodal diffusion core simulator to model the full reactor. Limitations in OSCAR-3 led to the need for improved modelling capabilities and better physics models for components present in the reactor core. OSCAR-4 offers improvements over OSCAR-3 in its approach to homogenization, and the new version of the diffusion core simulator allows for better modelling of movable components such as control rods. Fuel inventories calculated using OSCAR-4 can also easily be exported to MCNP, which allows the calculation of individual plate powers and local reaction rates amongst others. For these reasons OSCAR-4 is currently being introduced as a core support tool at the High Flux Reactor. In this work the steps that were followed to validate the reactor models are presented, and include results of validation calculations from both OSCAR-4 and MCNP6 over multiple reactor cycles. In addition differences in cross section library evaluations and their impact on the results are presented for the MCNP model.

HTR-PM fuel pebble irradiation qualification in the high flux reactor in Petten

NRG has performed a High Temperature Reactor fuel qualification test in the framework of the Chinese HTR development for INET – Tsinghua University, as a part of the qualification process of Chinese fuel for use in the HTR-PM (High Temperature Reactor-Pebble bed Module). Fuel qualification is an essential part of the licensing process of the HTR-PM, currently under construction in the Shandong province in the People’s Republic of China.The design of the INET irradiation setup for irradiation in the High Flux Reactor is performed by NRG. In the design of the irradiation rig, the pebbles are encased in graphite half shells to form a solid cylindrical stack. Within the graphite, thermocouples are located for temperature monitoring, and neutron fluence monitor sets for neutron fluence verification. Next to the temperature measurements, the online monitoring consisted of fission product release measurements. These measurements were performed with the Sweep Loop Facility, designed and built by JRC-IET.Post Irradiation Examinations are performed and consisted of gamma scanning, dimensional and weight measurements and visual inspection. In April 2016, the pebbles were transported from Petten to Karlsruhe, Germany where the second required step of fuel qualification will be performed, namely heating tests in the KÜFA facility.This paper describes the global lay-out and functionality of the irradiation facility. The online data generated and irradiation parameters are reported, with focus on the measured fission product release behavior of the fuel. The paper also presents non-destructive post irradiation examinations.

The MARINE experiment: Irradiation of sphere-pac fuel and pellets of UO2−x for americium breading blanket concept

Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like 241Am is therefore an option for the reduction of radiotoxicity and heat production of waste packages to be stored in a repository. The MARINE irradiation experiment is the latest of a series of European experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS, MARIOS, SPHERE) performed in the High Flux Reactor (HFR). The MARINE experiment is developed and carried out in the framework of the collaborative research project PELGRIMM of the EURATOM 7th Framework Programme (FP7). During the past years of experimental works in the field of transmutation and tests of innovative nuclear fuels, the release or trapping of helium as well as swelling have been shown to be the key issues for the design of such kind of fuel both as drivers and even more for Am-bearing blanket targets (due to the higher Am contents). The main objective of the MARINE experiment is to study the in-pile behaviour of uranium oxide fuel containing 13% of americium and to compare the behaviour of sphere-pac versus pellet fuel, in particular the role of microstructure and temperature on fission gas release and He on fuel swelling.The MARINE experiment will be irradiated in 2016 in the HFR in Petten (The Netherlands) and is expected to be completed in spring 2017.This paper discusses the rationale and objective of the MARINE experiment and provides a general description of its design for which some innovative features have been adopted.

The neutronics scheme adopted for the HELIOS irradiation experiment in the High Flux Reactor Petten

A dedicated neutronics computational scheme is a crucial tool in the preparation and the post-analysis of an irradiation experiment, as it defines the key parameters of the experiment. The complex and very dynamic environment of a Materials Testing Reactor (MTR) core, on the other hand, represents a challenging object for neutronics calculations. Within the framework of the HELIOS experiment conducted at the High Flux Reactor (HFR) in Petten (The Netherlands) in 2009 under the European Project EUROTRANS, a new neutronics scheme has been developed with the objective of finding the optimum balance between accuracy and calculation performance. This scheme is based on the Monte-Carlo technique and employs a two-level approach by separating the problem into overlapping spatial and temporal domains. The underlying physical phenomena are presented, together with elements of validation which gives a further insight into the effects governing the neutron fluxes in the core.

KüFA safety testing of HTR fuel pebbles irradiated in the High Flux Reactor in Petten

The Cold Finger Apparatus (KühlFinger-Apparatur—KüFA) in operation at JRC-ITU is designed to experimentally scrutinize the effects of Depressurization LOss of Forced Circulation (D-LOFC) accident scenarios on irradiated High Temperature Reactor (HTR) fuel pebbles. Up to 1600°C, the reference maximum temperature for these accidents, high-quality German HTR fuel pebbles have already demonstrated a small fission product release.This paper discusses and compares the releases obtained from KüFA-testing the pebbles HFR-K5/3 and HFR-EU1/3, which were both irradiated in the High Flux Reactor (HFR) in Petten. We present the time-dependent fractional release of the volatile fission product 137Cs as well as the fission gas 85Kr for both pebbles. For HFR-EU1/3 the isotopes 134Cs and 154Eu as well as the shorter-lived 110mAg have also been measured. A detailed description of the experimental setup and its accuracy is given. The data for the recently tested pebbles is discussed in the context of previous results.

SPHERE: Irradiation of sphere-pac fuel of UPuO2−x containing 3% Americium

Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like 241Am is therefore an option for the reduction of radiotoxicity of waste packages to be stored in a repository. The SPHERE irradiation experiment is the latest of a series of European experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS, MARIOS) performed in the HFR (High Flux Reactor). The SPHERE experiment is carried out in the framework of the 4-year project FAIRFUELS of the EURATOM 7th Framework Programme (FP7). During the past years of experimental works in the field of transmutation and tests of innovative nuclear fuels, the release or trapping of helium as well as helium induced fuel swelling have been shown to be the key issues for the design of Am-bearing targets. The main objective of the SPHERE experiment is to study the in-pile behaviour of fuel containing 3% of americium and to compare the behaviour of sphere-pac fuel to pellet fuel, in particular the role of microstructure and temperature on fission gas release (mainly He) and on fuel swelling.The SPHERE experiment is being irradiated since September 2013 in the HFR in Petten (The Netherlands) and is expected to be terminated in spring 2015. The experiment has been designed to last up to 18 reactor cycles (corresponding to 18 months) but may reach its target earlier.This paper discusses the rationale and objective of the SPHERE experiment and provides a general description of its design.

Overview of 10 years of irradiation projects on Eurofer97 steel at High Flux Reactor in Petten

The present paper demonstrates the overview of ten years NRG effort in investigating the Eurofer97 steel mechanical properties response after neutron irradiation in the High Flux Reactor in Petten, and highlights the main trends found during analyses of the extensive mechanical data sets.

Characteristics of microstructure, swelling and mechanical behaviour of titanium beryllide samples after high-dose neutron irradiation at 740 and 873 K

Titanium beryllide Be12Ti is one of the candidate materials for advanced neutron multiplier in the helium-cooled breeding blanket of Demonstration Power Plant (DEMO).Cylinder-shaped specimens (diameters of 3mm) consisting, mainly, of Be12Ti phase were investigated after high-dose neutron irradiation at 740 and 873K in High Flux Reactor in Petten, Netherlands.The sizes and shapes of radiation-induced gas bubbles as well as swelling of samples were evaluated by means of transmission electron microscopy (TEM). Mechanical behaviour of titanium beryllide specimens in initial and irradiated states was investigated using compression tests under constant loading.Obtained results are discussed and compared with available data on pure beryllium.

Tritium permeation and desorption in reduced activation martensitic steels studied in EXOTIC-9/1 irradiation experiment

The EXOTIC-9/1 has been irradiated in the High Flux Reactor in Petten within the European framework for the development of the helium cooled pebble bed breeder concept. The EXOTIC-9/1 irradiation assembly consists of a pebble bed with Li2TiO3 pebbles enclosed in a EUROFER97 containment. The irradiation of EXOTIC-9/1 proceeded for 301 full power days reaching 1.3dpa in steel. Irradiation temperatures varied between 613 and 853K.Tritium permeation in a EUROFER97 containment wall was studied, in-pile, during temperature and purge gas chemistry transients. Under reference purge gas conditions (He+0.1%H2) permeation proceeded in the diffusion-limited regime. Tritium permeability in EUROFER97 was obtained from analysis of the in-pile data.During the post irradiation examination program a number of EUROFER97 containment wall segments were ramp annealed in a temperature programmed desorption setup. This study determines the amount of tritium retained in the samples and presents analysis of the desorption spectra.

Analysis of tritium retention in beryllium pebbles in EXOTIC, PBA and HIDOBE-01 experiments

Tritium desorption studies were performed on beryllium pebbles irradiated in three different irradiation campaigns (i.e., EXOTIC-8, PBA and HIDOBE-01), which were carried out in the High Flux Reactor in Petten. Irradiation doses corresponded to the following helium production levels: 100 appm in EXOTIC-8/3, 284 appm in EXOTIC-8/7, 260 appm in EXOTIC-8/8, 400–500 appm in PBA, and 3090 appm in HIDOBE-01. The beryllium pebbles used in this study were manufactured by melt spray and rotating electrode methods. Measured tritium inventories were compared with the MCNP/FISPACT calculations, and it was concluded that for the irradiation temperatures below 900K the tritium retention is nearly 100% of tritium production. At the irradiation temperatures above 1000K, tritium escaped from beryllium during irradiation and the retention fraction, in case of the HIDOBE-01 experiment, was about 5% or less.Analysis of the release spectra revealed a number of characteristic peaks measured as gas (HT) and water (HTO). Below 500K, tritium release takes place in the HTO form and can be ascribed to the exchange processes taking place at the beryllium surface and tritium dissociation from small clusters. At higher temperatures, tritium release takes place in the HT form and in most spectra represented as one peak centered between 1100 and 1200K.Optical and scanning electron microscopy studies demonstrated that between 1073 and 1173K, helium bubbles form an extended network of interconnected channels, which are apparently responsible for the observed tritium release.

Lithium orthosilicate surfaces: Characterization and effect on tritium release

Within the European Union, slightly hyperstoichiometric lithium orthosilicate has evolved as one of the candidate solid breeder materials for the helium cooled pebble bed blanket, which will be tested in ITER. In the past, several long-term irradiation experiments proved that lithium orthosilicate shows excellent tritium release behavior when purged with helium with 0.1% hydrogen.In this work, short-term irradiation experiments at the High Flux Reactor in Petten with two standard pebble qualities, as-received and heat-treated lithium orthosilicate were carried out to investigate possible differences in tritium release. Since the surface of the pebbles may play a significant role, especially for short-term irradiation, and the mechanism of desorption from the surface determines the chemical form and very likely the release rate of tritium, the aim of this work is to link the chemical composition of the surface to the Thermal Programmed Desorption (TPD) tritium release experiments. Therefore X-ray photoelectron spectroscopy and time-of-flight secondary ion mass spectrometry were applied to characterize the surface of the unirradiated pebbles and to obtain sputter depth profiles of up to 800nm. In addition, X-ray diffractometry was used to determine the composition of phases of the samples before and after irradiation.

PYCASSO: Irradiation of HTR coated particles at high temperatures

The PYCASSO (PYrocarbon irradiation for Creep And Swelling/Shrinkage of Objects) irradiations were a part of RAPHAEL (ReActor for Process Heat And Electricity), a 4-year Integrated Project on Very High Temperature Reactors, which started in April 2005 as part of the European 6th Framework Program. In the Fuel Technology Sub-Project (SP-FT), the PYCASSO experiments were performed to determine the effect of high temperature (900–1100°C) irradiation on thermo-mechanical properties of various coating materials for TRISO coated particle fuel. Effects due to the presence of fuel, such as pressurization or chemical attack by fission products, have been excluded by irradiating surrogate (ZrO2 and Al2O3) kernels. The partners involved in this irradiation are CEA (France), JAEA (Japan) and KAERI (Republic of South Korea). The PYCASSO irradiations have taken place in the High Flux Reactor (HFR) Petten and were coordinated by NRG (The Netherlands). In this paper we will discuss the performance of PYCASSO-I by a direct comparison between the in-pile behavior of the experiment and the behavior of the corresponding finite element model. The experience of PYCASSO-I led to recommendations for the follow-up irradiation PYCASSO-II irradiation, of which the performance is also presented. The PYCASSO-I experiment has been successfully dismantled, which was a complex procedure of retrieving samples from the highly activated sample holders. The dismantling was successful, as demonstrated by relatively low sample activity, which will most likely allow exporting the samples from hot cells and enable PIE in a glovebox. Also the measured fluence is reported, which shows that the experiment fluence target has been perfectly met. Finally, the proposed PIE for the PYCASSO-I and PYCASSO-II samples in the ARCHER FP7 proposal is presented.

JRC's on-line fission gas release monitoring system in the high flux reactor Petten

For HTR fuel irradiation tests in the HFR Petten a specific installation was designed and installed, dubbed the “Sweep Loop Facility” (SLF). The SLF is tasked with three functions, namely temperature control by gas mixture technique, surveillance of safety parameters (temperature, pressure, radioactivity etc.) and analysis of fission gas release for three individual capsules in two separate experiments. The SLF enables continuous and independent surveillance of all gas circuits. The release of volatile fission products of the in-pile experiments is monitored by continuous gas purging. The fractional release of these fission products, defined as the ratio between release rate of a gaseous fission isotope (measured) to its instantaneous birth rate (calculated), is a licensing-relevant test for HTR fuel. The newly developed gamma spectrometry station allows for higher measurement frequencies, thus enabling follow-up of rapid and massive release transients. The designed stand-alone system was tested and fully used through the final irradiation period of the HFR-EU1 experiment which was terminated on 18 February 2010. Its robustness allowed us to use it as extra safety instrumentation. This paper describes the gas activity measurement technique based on HPGe gamma spectrometry and illustrates how qualitative and quantitative analysis of volatile fission products can be performed on-line.

New Temperature Monitoring Devices for High-Temperature Irradiation Experiments in the High Flux Reactor Petten

Within the European High Temperature Reactor Technology Network (HTR-TN) and related projects a number of HTR fuel irradiations are planned in the High Flux Reactor Petten (HFR), The Netherlands, with the objective to explore the potential of recently produced fuel for even higher temperature and burn-up. Irradiating fuel under defined conditions to extremely high burn-ups will provide a better understanding of fission product release and failure mechanisms if particle failure occurs. After an overview of the irradiation rigs used in the HFR, this paper sums up data collected from previous irradiation tests in terms of thermocouple data. Some R&D for further improvement of thermocouples and other on-line instrumentation will be outlined.

Irradiation response of ODS Eurofer97 steel

Oxide dispersion strengthened (ODS) Eurofer97 steel (EU batch, 0.3wt.% of Y2O3 particles), produced by mechanical alloying followed by hot rolling, is irradiated in the High Flux Reactor in Petten, The Netherlands at three different irradiation temperatures (300, 450 and 550°C) up to nominal doses of 1dpa and 3dpa. The effect of neutron irradiation on the mechanical properties of ODS Eurofer97 material is investigated. It is shown that the irradiation hardening of ODS Eurofer97 steel occurs at 300°C, whereas during irradiation at 450 and 550°C no changes in mechanical properties are observed compared to the unirradiated material. This effect is possibly a result of the annealing of the irradiation damage at temperatures higher than 300°C. The observed shifts in the Ductile to Brittle Transition Temperatures due to irradiation at different temperatures are discussed and compared with non-ODS Eurofer97 steel.

Accident testing of high-temperature reactor fuel elements from the HFR-EU1bis irradiation

Four spherical high-temperature reactor fuel elements irradiated at the High Flux Reactor in Petten within the HFR-EU1bis irradiation campaign were transported to JRC-ITU for post-irradiation examination and accident testing. This article reports the accident tests performed on the first two fuel elements HFR-EU1bis/1 and HFR-EU1bis/3 with the Küfa device. The fuel elements were heated up to the irradiation temperature (1250 °C) and then to accident temperatures (1600 °C and 1700 °C, respectively) for durations of several hundred hours to simulate depressurization and loss-of-forced-circulation accidents (D-LOFC). The overall low fractional release of 85Kr and the absence of steps in the release curves suggest that no coated particles failed during the accident simulations. On the other hand, the fractional release of 134Cs and 137Cs was much higher than in previous experiments, indicating increased diffusion through the coating layers at the higher irradiation temperature, as confirmed with a finite-difference diffusion calculation. Finally, the calculation suggests that the high fractional silver and cesium releases at 1250 °C are due to a matrix contamination prior to the Küfa accident testing.