High Flux Reactor Review and Reactivity Control Analysis

For over 60 years, research reactors have provided the world with a versatile tool to test materials and promote irradiation research, as well as to produce radioisotopes for medical treatments. The High Flux Reactor (HFR), as a water moderated and cooled, beryllium-reflected reactor has awarded more attention in recent years. There is a wide range of designs and applications for HFRs that based on their own situation to meet research requirements. For the purpose of reducing the volume and mass of the reactor, as well as ensuring the safety operation, it is necessary to determine the most effective reactivity control scheme, and analyze the corresponding reactivity insertion accidents. This paper is going to investigate typical high flux reactors within the same type with HFETR, summarize general description and characteristics, the uses of the high flux reactor, and reactivity control mechanisms. In addition, the associated reactivity insertion accidents were presented and analyzed. The analysis result will provide some references to further design and construction of high flux reactor.

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.

Modelling of the H2020 INSPYRE Fuel Creep Experiment

To better understand irradiation creep of nuclear fuel, NRG has prepared, as part of the H2020 European project INSPYRE, a separate effect irradiation experiment in the High Flux Reactor (HFR) in Petten (the Netherlands) aiming to measure fuel creep in-pile as a function of temperature, flux, burn-up and axial pressure load. This continuous type of measurement will supply a large data set, leading to more detailed knowledge on fuel behaviour during irradiation. To support the experiment and make optimal use of the generated data, a model is created of the experiment to better predict the behaviour of the fuel samples during irradiation. The current paper describes the numerical model, which couples the 1.5D fuel performance code TRANSURANUS (TU) with a Finite Element Analysis (FEA). The thermal analysis of the experiment is carried out using the FEA. Such approach enables to model a rather complex geometry of the experiment, and to include axial heat transport, which is not implemented in TU. TU is modified in order to use the fuel pellet temperatures obtained using the FEA and to include the axial load present in the experiment. The model is validated against several test cases and used to predict the fuel behaviour during a selection of foreseen irradiation scenario’s. Results of the model will be used in the future for optimization of the irradiation parameters used in the experiment and for analysis of the data obtained during the irradiation.

Research on the Seismic Analysis and Optimization of High Flux Reactor Nuclear Power Piping System Based on Multi-Point Response Spectrum Method

High flux reactor is an important engineering test reactor, which can be used in irradiation research of materials, chemistry, isotopes, medicine and other fields. In the high flux reactor coolant system, there are a large number of nuclear pipes and the layout is complex. The optimization of seismic analysis method for reactor coolant system is an important part in the design process to ensure the nuclear pipes meet the design specifications. The traditional single point response spectrum method needs to envelope the response spectrum of different floors as the analysis input. This method is difficult to give the reasonable seismic load to the numerous nuclear pipes and it will increase the design cost and the difficulty of safety analysis about nuclear pipe. In this paper, an optimized seismic analysis method of reactor coolant system is proposed. By using the multi-point response spectrum method, the optimization of different excitation loading modes for different constrained support points is realized. The analysis results show that the multi-point response spectrum method can solve the problem that different support points are located at different elevation floors in the reactor coolant system, which makes the calculation results more accurate and reasonable. Compared with the traditional method, it can make the design more efficient and practical.

High Flux Reactor Continued Safe Operation: Time Limited Ageing Analyses

The High Flux Reactor (HFR) is a multipurpose nuclear reactor located in Petten, the Netherlands. With its 45 MWth it is one of the most powerful and versatile research reactors in the world. Its main roles are material irradiation and medical isotopes production. The output of the reactor in terms of medical isotopes is important at a global level (60% of European demand). Every day in the Netherlands alone 30.000 patients are treated using isotopes produced in the HFR. The importance of the HFR dates back in time. The HFR has been in service since 1961. Due to the long life (58 years to date) of the reactor an efficient integrated ageing management program (AMP) is envisaged as it is foreseen that the HFR will continue to operate for a prolonged period of time. The development of the AMP has begun in 2018 (CSO project) and will be completed in view of the IAEA CSO mission. The HFR is the second reactor in the world to undergo this type of IAEA review and one of the objectives of this project is to set a state of the art when it comes to research reactors long term operation. The CSO project foresees four major sections: scoping and screening, development and improvement of plant programs, (re) validation of time limited ageing analyses (TLAAs) and realization of the ageing management program. In this paper the focus will lie on the TLAAs. The applicable TLAAs were scoped starting from the IGALLs TLAAs list. The TLAAs relevant for the HFR are: TLAA fatigue, TLAA reactor vessel, TLAA leak before break, TLAA manufacturing flaws TLAA beryllium and TLAA equipment qualification. The latter was developed in the framework of the equipment qualification plant program and does not figure as an independent TLAA in the CSO project. For each TLAA the principal problematics will be highlighted and the possible solutions illustrated.

An Update on the Investigation of Fracture Toughness Properties of the High Flux Reactor Vessel From Surveillance Test Campaign in 2017

The reactor vessel of the High Flux Reactor (HFR) in Petten has been fabricated from Al 5154-O alloy grade with a maximum Mg content of 3.5 wt. %. The vessel experiences large amount of neutron fluences (notably at hot spot), of the order of 1027 n/m2, during its operational life. Substantial damage to the material’s microstructure and mechanical properties can occur at these high fluence conditions. To this end, a dedicated surveillance program: SURP (SURveillance Program) is executed to understand, predict and measure the influence of neutron radiation damage on the mechanical properties of the vessel material. In the SURP program, test specimens fabricated from representative HFR vessel material are continuously irradiated in two specially designed experimental rigs. A number of surveillance specimens are periodically extracted and tested to evaluate the changes in fracture toughness properties of the vessel as function neutron fluence. The surveillance testing results of test campaigns performed until 2015 were published previously in [1, 2]. The current paper presents fracture toughness and SEM results from the recent surveillance campaign performed in 2017. The fracture toughness specimen tested in this campaign received a thermal neutron fluence of 13.56 x1026 n/m2, which is ∼8.9 × 1025 n/m2 more than the thermal fluence received by the specimen tested in SURP 2015 campaign. These results from this campaign have shown no change in the fracture toughness from the values measured in the previous SURP campaign. The SEM observations are performed to study the fracture surface, to measure (by WDS) the transmutation Si formed near crack tip and to investigate various inclusions in the microstructure. SEM fracture surface investigation revealed a tortuous (bumpy) fracture surface constituting micro-scale dimples over majority of the fracture area. Islands of cleavage facets and secondary cracks have been observed as well. EDS analysis of various inclusions in the microstructure revealed presence of Fe rich inclusions and Mg-Si rich precipitates. Additionally, inclusions rich in Al-Mg-Cr-Ti were identified. Finally, changes in mechanical properties of Al 5154-O alloy with an increase in neutron fluence (or transmutation Si) are discussed in correlation with SEM microstructure and fracture morphology observed in SEM. TEM investigation of precipitate microstructure is ongoing and those results will be published in future.

Continued Safe Operation (LTO Research Reactors) High Flux Reactor, Petten

The HFR is one of the most powerful and versatile Research Reactors in the world. Because of its strategical importance in the medical isotopes production, after 57 years of operating experience a Continued Safe Operations (CSO) mission will take place. The CSO project structure is based on the outline given in the IAEA draft Safety Guide SSG-48. The ageing management for the HFR is evaluated based on the IAEA Specific Safety Guide SSG-10. Moreover, the approach to the contents of the project is supported by the IAEA Draft Guidelines for Peer Review of “Ageing Management of Research Reactors for Continued Safe Operations”. The HFR is the second Research Reactor (RR) in the world to undergo this type of assessment and its experience will be extremely valuable in setting the international standards for CSO of research reactors. This paper describes the phases of the CSO project, the challenges encountered and the experience built during its development.