HFR Petten Research Articles

Investigation of Irradiation Hardening and Effectiveness of Post-Irradiation Annealing on the Recovery of Tensile Properties of VVER-1000 Realistic Welds Irradiated in the LYRA-10 Experiment

Assessing the embrittlement and hardening of reactor pressure vessel steels is critical for the extension of the service lifetime of nuclear power plants. This paper summarises the tensile test results on the irradiation behaviour of realistic VVER-1000 welds from the STRUMAT-LTO project. The welds were irradiated at the HFR (Petten, the Netherlands) to a fluence of up to 1.087 × 1020 n·cm−2, and their irradiation hardening was studied by means of tensile testing. The four grades, with different Mn and Ni contents, show different hardening behaviours. The highest degree of irradiation hardening is observed for the weld that has the highest combined Ni + Mn content. The results show that there is a synergetic effect of Mn and Ni on the irradiation hardening behaviour of the VVER-1000 welds. Besides irradiation hardening, the effectiveness of post-irradiation annealing treatments on the recovery of the tensile properties is studied in the present work. Post-irradiation annealing treatments conducted at 418 °C and at 475 °C proved to be effective for three of the four investigated welds. For the realistic weld with the highest combined Ni + Mn, only the annealing at 475 °C led to the complete recovery of the tensile properties.

Irradiation of thorium-bearing molten fluoride salt in graphite crucibles

Four fluoride fuel salt samples (78LiF-22ThF4) in graphite crucibles were irradiated in the HFR Petten for a duration of 508 Full Power Days under the name SALIENT-01 (SALt Irradiation ExperimeNT). Goal of the experiment was to gain experience with the design of liquid salt experiments and the handling of the salts before and after irradiation. Specific research goals for SALIENT-01 are (i) to confirm claims of good fission product retention in the salt, (ii) to obtain size distributions for noble metal particles using Transmission Electron Microscopy and (iii) to assess possible interactions between fuel salt and fine-grained nuclear graphite, as well as possible uptake of fission products by the graphite. Here the design and irradiation history of the experiment are discussed together with plans for post-irradiation examinations. Limitations in representativeness of this experiment and capsule irradiations in general are discussed as well as follow-up actions to improve the quality of future irradiations.

Burn-up determination and accident testing of HTR-PM fuel elements irradiated in the HFR Petten

The current paper reports the results of gamma spectroscopic burn-up determination and KüFA safety testing at JRC Karlsruhe on spherical high-temperature reactor fuel elements, which were fabricated by the Institute of Nuclear and New Energy Technology of the Tsinghua University, Beijing. The fuel elements were irradiated in the High Flux Reactor, Petten, in the frame of the HFR-EU1 and HTR-PM campaigns and transported to JRC Karlsruhe for post-irradiation examination and accident testing in the KüFA device in the frame of a bilateral safety research study. Burn-up determination was performed on seven fuel elements based on the quantitative measurement of their Cs-137 inventories using an established gamma spectroscopy set-up in the JRC hot cell facilities. Accident testing was conducted on three HTR-PM fuel elements in several phases at simulated accident temperatures between 1620 °C and 1770 °C for 150 h to simulate hypothetical depressurization and loss-of-forced circulation accidents. The release of Kr-85 was measured during the tests in a cold trap and solid fission product release was determined by gamma spectroscopic analyses of exchangeable cold plates in a low background environment. After successful completion of the KüFA tests the fuel elements are currently undergoing further post-irradiation examinations including profile disintegration as well as ceramography and scanning electron microscopy of individual coated particles.

Irradiation and post-irradiation examination of uranium-free nitride fuel

Two identical Phénix-type 15-15Ti steel pinlets each containing a 70 mm Pu0.3Zr0.7N fuel stack in a 1-bar helium atmosphere have been irradiated in the HFR Petten at medium high linear power (46–47 kW/m at BOL) and an average cladding temperature of 505 °C. The pins were irradiated to a plutonium burn-up of 9.7% (88 MWd/kgHM) in 170 full power days. Both pins remained fully intact. Post-irradiation examination performed at NRG and PSI showed that the overall swelling rate of the fuel was 0.92 vol-%/%FIHMA. Fission gas release was 5–6%, while helium release was larger than 50%. No fuel restructuring was observed, and only mild cracking. EPMA measurements show a burn-up increase toward the pellet edge of up to 4 times. All investigated fission products except to some extent the noble metals were found to be evenly distributed over the matrix, indicating good solubility. Local formation of a secondary phase with high Pu content and hardly any Zr was observed. A general conclusion of this investigation is that ZrN is a suitable inert matrix for burning plutonium at high destruction rates.

Tritium release from highly neutron irradiated constrained and unconstrained beryllium pebbles

Beryllium is the reference neutron multiplier material in the Helium Cooled Pebble Bed (HCPB) breeding blanket of fusion power plants. Significant tritium inventory accumulated in beryllium as a result of neutron-induced transmutations could become a safety issue for the operation of such blankets as well as for the nuclear waste utilization. To provide a related materials database, a neutron irradiation campaign of beryllium pebbles with diameters of 0.5 and 1mm at 686–1006K, the HIDOBE-01 experiment, has been performed in the HFR in Petten, the Netherlands, producing up to 3020appm helium and 298appm tritium. Thermal desorption tests of irradiated unconstrained and constrained beryllium pebbles were performed in a purge gas flow using a quadrupole mass-spectrometer (QMS) and an ionization chamber. Compared to unconstrained pebbles, constrained beryllium pebbles have an enhanced tritium release at all temperatures investigated. Small elongated sub-grains formed under irradiation in the constrained pebbles promote formation of numerous channels for facilitated tritium release.

Mechanical compression tests of beryllium pebbles after neutron irradiation up to 3000 appm helium production

Resultsof mechanical compression tests of irradiated and non-irradiated beryllium pebbles with diameters of 1 and 2mm are presented. The neutron irradiation was performed in the HFR in Petten, The Netherlands at 686–968K up to 1890–2950appm helium production. The irradiation at 686 and 753K cause irradiation hardening due to the gas bubble formation in beryllium. The irradiation-induced hardening leads to decrease of steady-state strain-rates of irradiated beryllium pebbles compared to non-irradiated ones. In contrary, after irradiation at higher temperatures of 861 and 968K, the steady-state strain-rates of the pebbles increase because annealing of irradiation defects and softening of the material take place. It was shown that the steady-state strain-rates of irradiated beryllium pebbles always exceed their swelling rates.

Tritium release and retention properties of highly neutron-irradiated beryllium pebbles from HIDOBE-01 experiment

The current helium cooled pebble bed (HCPB) tritium breeding blanket concept for fusion reactors includes a bed of 1mm diameter beryllium pebbles to act as a neutron multiplier. Beryllium pebbles, fabricated by the rotating electrode method, were neutron irradiated in the HFR in Petten within the HIDOBE-01 experiment. This study presents tritium release and retention properties and data on microstructure evolution of beryllium pebbles irradiated at 630, 740, 873, 948K up to a damage dose of 18dpa, corresponding to a helium accumulation of about 3000appm. The measured cumulative released activity from the beryllium pebbles irradiated at 948K was found to be significantly lower than the calculated value. After irradiation at 873 and 948K scanning electron microscopy (SEM) and transmission electron microscopy (TEM) analyses revealed large pores or bubbles in the bulk and oxide films with a thickness of up to 8μm at the surface of the beryllium pebbles. The radiation-enhanced diffusion of tritium and the formation of open porosity networks accelerate the tritium release from the beryllium pebbles during the high-flux neutron irradiation.

Results of the HFR-EU1 fuel irradiation of INET and AVR pebbles in the HFR Petten

The HFR-EU1 irradiation in the HFR Petten was successfully conducted between 2006 and 2010 by the European Commission's Joint Research Centre, Institute for Energy and Transport (JRC-IET). Two recently produced INET fuel pebbles and 3 historical AVR pebbles, both containing TRISO fuel particles, were irradiated at representative temperatures to burn-ups up to 14.3% FIMA, to be confirmed by measurements. The objective was to test fuel beyond the specifications of current pebble bed cores to demonstrate that HTRs are capable of enhanced sustainability (further increased power conversion efficiency, improved fuel use) and thus reduced waste production. INET and AVR pebbles were irradiated and monitored in separately controlled capsules of the same irradiation rig. Fission gas release was monitored by HP Ge gamma spectrometry thus enabling evaluation of the characteristic release over birth fraction which is indicative for the health of the fuel. In none of the pebbles, abnormally increased fission gas release was observed indicating that all of the approx. 45,000 coated particles in the pebbles had remained intact.Unexpectedly high thermocouple failure in one of the capsules had delayed the experiment because it required construction of a new safety case. This paper presents the irradiation history of the experiment including fission gas release data.

Comparison of two irradiation testing results of HTR-10 fuel spheres

In order to study the irradiation performance of fuel spheres for the 10MW high-temperature gas-cooled reactor (HTR-10), two in-pile irradiation tests were performed.The first irradiation test was performed in the Russian reactor IVV-2M from July 2000 to February 2003. The four irradiated fuel spheres were randomly taken from the first product batches. The central temperature of irradiated fuel spheres during this irradiation was kept between 950 and 1050°C. The maximum burnup of the irradiated spheres was 107,000MWd/tU and the maximum fast neutron fluence (E>0.1MeV) was 1.31×1025m−2. The release-to-birth rate ratios (R/B) of Kr and Xe fission products were 10−7 to 10−5.The second irradiation test was performed in HFR Petten (HFR-EU1) from September 29, 2006 to February 19, 2010. Experiment HFR-EU1 consisted of 3 German AVR pebbles and 2 INET pebbles. The two INET pebbles were randomly taken from the latest production batch. The surface temperature of pebble INET 2 during irradiation was approx. 940°C. The experiment was performed in two campaigns from 29 September 2006 to 24 February 2008 (12 reactor cycles of 28 days) and continued from 19 October 2009 to 19 February 2010 (4 reactor cycles). Up to end of irradiation, the experiment had accumulated 16 reactor cycles totaling 445 efpd and calculated burnup and maximum fast neutron fluence (E>0.1MeV) were 9.3% (pebble INET 1) and 11.6%FIMA (pebble INET 2) and about 4.95×1025m−2, respectively. At end of irradiation, based on 85mKr, R/B of approx. 8×10−8 was measured.The second irradiation test results are much better than the first. The irradiated sample manufacture technology and characterization, irradiation conditions and results of two irradiation tests are compared in detail in this paper.

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.

Determination of boron concentration in blood and tissue samples from patients with liver metastases of colorectal carcinoma using Prompt Gamma Ray Activation Analysis (PGAA)

As part of the studies on Boron Neutron Capture Therapy at the University of Mainz, Germany, a clinical trial has been started in which, four patients suffering from liver metastases of colorectal carcinoma have been enrolled. Specimens of blood and healthy tissue samples taken from the patients were measured at the PGAA facilities at the HFR in Petten, The Netherlands, and at the FRM II in Munich, Germany. From the measured boron concentrations, pharmacokinetic curves and blood-to-tissue concentration ratios were produced.

Neutron Dosimetry and 3-D Neutron Transport Calculations as Toolkit at HFR Petten for Detailed Neutron Monitoring and Damage Analysis

Dosimetry at the High Flux Reactor in Petten, the Netherlands, serves a variety of purposes ranging from demonstrating compliance with the license for operating the reactor to validating the predictions of displacements per atom (DPA) in material irradiations. It is also used in the validation of the 3-D neutron transport calculations during post-irradiation analysis to normalize the spectrum obtained from these calculations. In this paper, a thorough examination of irradiated steel specimens is carried out using the measurements performed with the neutron activation technique as well as the 3-D neutron transport calculations to check the consistency in the resulting DPA values.

Out of pile tritium release behaviour and microscopic investigation of lithium metatitanate irradiated in the High Flux Reactor in Petten

The irradiation experiment EXOTIC 9/1 is the latest in a series of EXOTICs that has been irradiated in the high flux reactor in Petten and studies tritium release and inventory in ceramic pebbles containing lithium, which are key properties for the fusion fuel cycle. New production routes of pebbles are developed, leading to different thermomechanical and tritium release properties. The pebbles, produced by an extrusion–spheroidisation–sintering process at CEA in 2004, differ in porosity and density from the Li 2TiO 3 ceramics tested in the previous EXOTIC 8 programme. The pebbles have a diameter in the range from 0.6 to 0.8 mm, a high density (93% TD) and an open porosity of 1.7%. The pebbles mainly differ in enrichment (7.5% and 30% 6Li-enrichment). During irradiation of 300 Full Power Days in the HFR in Petten, the temperature in EXOTIC 9/1 was varied between 340 and 580 °C. The final irradiation (EOL) temperature is 480 °C and the average total 6Li burn-up is 3%. The total measured activity from tritium during irradiation is 220.42 Ci. After the completion of the irradiation, the out of pile release behaviour was studied by temperature programmed desorption (TPD). More specifically, the remaining tritium inventory in the pebbles and the activation energy of different release mechanisms were studied. The TPD was carried out for Li 2TiO 3 samples with 7.5% (natural) and 30% (enriched) 6Li-abundance and revealed very different curves, indicating that the materials have different release mechanisms. Dissolving the irradiated samples in hydrogen peroxide indicates that no tritium is left in the pebbles after TPD. Results from optical microscopy show it is likely that more intense bubble formation has taken place in the enriched samples, compared to the natural samples.

Modelling of the HFR-EU1BIS experiment and thermomechanical evaluation

Numerical modelling has been successfully applied during the design and evaluation of the HFREU1bis HTR fuel pebble irradiation experiment conducted by the Joint Research Centre Institute for Energy (JRC) in the HFR Petten, The Netherlands. HFR-EU1bis contains 5 HTR fuel pebbles that are placed in a graphite shroud. This assembly is placed in a containment tube, which is again placed in another containment, which is in contact with the HFR cooling water. The experimental requirement is to maintain a central temperature of 1250 °C for all pebbles throughout the irradiation time of 250 effective full power days (efpd). This was achieved by tailoring the gas gaps in the sample holder such that the axial heat generation profile is compensated. Determining the dimensions of the components of the experiment has been done using a thermomechanical finite element model. A combination of numerical and analytical techniques has been applied to minimise calculation times. The thermomechanical finite element model includes the thermal influence of changing gas gap dimensions due to expansion. Heat transfer by radiation through the gas gap has been modelled as well. Results of the model have been compared with measured temperatures. After some adjustments of the model parameters, good agreement has been found between calculated and measured temperatures. Additionally, the influence of irradiation on graphite thermal conductivity and thermal expansion has been included in the model. Due to irradiation damage the thermal conductivity of graphite decreases and thermal expansion generally increases, leading to increasing thermal stresses. Due to burn-up however, the heat flux decreases during irradiation, which leads to a decrease in thermal stresses. To evaluate the influence of these competing mechanisms, the thermal stress evolution during irradiation has been calculated and will be discussed.

Results of AVR fuel pebble irradiation at increased temperature and burn-up in the HFR Petten

The irradiation experiment HFR-EU1bis was performed by the European Commission's Joint Research Centre-Institute for Energy (JRC-IE) in the HFR Petten to test five spherical High Temperature Reactor (HTR) fuel pebbles of former German production with TRISO coated particles for their potential for very high temperature performance and high burn-up. The irradiation started on 9 September 2004 and was terminated on 18 October 2005 after 10 reactor cycles totaling 249 efpd and a maximum burn-up of 11.07% FIMA. The objective of the HFR-EU1bis test was to irradiate five HTR fuel pebbles at conditions beyond the characteristics of current HTR reactor designs with pebble bed cores, e.g. HTR-Modul, HTR-10 and PMBR. This should demonstrate that pebble bed HTRs are capable of enhanced performance in terms of sustainability (further increased power conversion efficiency, better use of fuel) and thus reduced waste production. The central temperature of all pebbles was kept as closely as possible at 1250 °C and held constant during the entire irradiation, with the exception of HFR downtime and power transients. This is the expected maximum central fuel temperature of a pebble bed VHTR with a coolant outlet temperature of 1000 °C. HFR-EU1bis should demonstrate the feasibility of low coated particle failure fractions under normal operating conditions and more specifically: • increased central fuel temperature of 1250 °C compared to 1000–1200 °C in earlier irradiation tests; • irradiation to a burn-up close to 16% FIMA, which is double the license limit of the HTR-Modul; due to a neutronics data processing error, the experiment was prematurely terminated at 11.07% FIMA maximum so that this objective was not fully achieved; • confirmation of low coated particle failure fractions due to temperature, burn-up and neutron fluence. This paper provides the irradiation history of the experiment including data on fission gas release. Post-irradiation examinations at NRG Petten and JRC-ITU Karlsruhe included the verification of the received neutron fluences, burn-up and spectrum. They will be followed shortly by safety-relevant heating tests at JRC-ITU to verify fission product retention by out-of-pile heating tests beyond 1600 °C.

Design of a Rotating Facility for Extracorporal Treatment of an Explanted Liver with Disseminated Metastases by Boron Neutron Capture Therapy with an Epithermal Neutron Beam

In 2001, at the TRIGA reactor of the University of Pavia (Italy), a patient suffering from diffuse liver metastases from an adenocarcinoma of the sigmoid was successfully treated by boron neutron capture therapy (BNCT). The procedure involved boron infusion prior to hepatectomy, irradiation of the explanted liver at the thermal column of the reactor, and subsequent reimplantation. A complete response was observed. This encouraging outcome stimulated the Essen/Petten BNCT group to investigate whether such an extracorporal irradiation could be performed at the BNCT irradiation facility at the HFR Petten (The Netherlands), which has very different irradiation characteristics than the Pavia facility. A computational study has been carried out. A rotating PMMA container with a liver, surrounded by PMMA and graphite, is simulated using the Monte Carlo code MCNP. Due to the rotation and neutron moderation of the PMMA container, the initial epithermal neutron beam provides a nearly homogeneous thermal neutron field in the liver. The main conditions for treatment as reported from the Pavia experiment, i.e. a thermal neutron fluence of 4 x 10(12) +/- 20% cm(-2), can be closely met at the HFR in an acceptable time, which, depending on the defined conditions, is between 140 and 180 min.

Extra-Corporal Treatment of Liver Metastases by BNCT at the HFR Petten

A computational investigation has shown that the epithermal neutron beam presently used for BNCT of the brain in the HFR in Petten The Netherlands can also be used for BNCT of extra-corporal livers. Liver BNCT has already been performed in Pavia Italy providing very promising results. By rotating the liver and surrounding it by Polymethylmethacrylate and graphite and with the appropriate dimensions of the set-up, the same homogeneity in the thermal neutron field in the liver can be obtained in Petten as in Pavia.

The HFR Petten high dose irradiation programme of beryllium for blanket application

This paper reports the objectives of a high dose irradiation programme of beryllium in the high flux reactor in Petten that will be performed in the frame of the European Programme for development of the helium cooled pebble bed (HCPB). The main objectives of irradiation are to study the irradiation behaviour of beryllium pebble beds, in terms of dimensional stability and their tritium retention behaviour. High fluence irradiation of beryllium specimens (HIDOBE) will be performed up to He contents of 6000 appm, which is about 30% of the expected DEMO end-of-life value with a relevant dPa to He ratio. In this paper, the nuclear parameters, irradiation conditions and the test-matrix, i.e. beryllium grades and pebbles are presented.

Irradiation Testing of Blanket Materials at the HFR Petten with on Line Tritium Monitoring

Irradiation experiments are performed in support of fusion blanket technology development. These comprise ceramic solid breeder materials, and a liquid Lithium Lead alloy, as well as blanket subassemblies and components. Experimental facilities at the HFR to study tritium release, permeation characteristics, and neutron irradiation performance, have recently been extended. This paper gives an overview on the tritium breeding materials irradiation programme and describes the facilities required for irradiation testing and on-line tritium measurement.

Irradiation of SiC f/SiC composites, fibres and matrices in HFR Petten

This paper reports on the pre-irradiation tests of SiC f/SiC materials as a contribution to the EFDA Technology Programme on advanced materials. A high temperature irradiation of SiC materials in the High flux Reactor in Petten, denoted as ‘SICCROWD’ has recently started. The SiC f/SiC materials included are supplied by the EU partners. Further materials are included in the frame of an IEA collaborative effort, by Japanese and USA partners. The irradiation matrix comprises a variety of fibres and matrix materials, as well as composites with 1-D, 2-D or 3-D architectures. CVD monolithic SiC serves as a reference material and should enable comparison with other, e.g. complementary, irradiation data. The specimen types are bars, for four-point bend tests, and discs for flash diffusivity measurements. Both types will be used to measure dimensional changes. Some specimens have been mounted to allow for in-situ determination of thermal conductivity degradation by neutron irradiation. The irradiation rig, materials and test matrix will be presented as well as the results of the pre-irradiation characterisations.