Periodic Reporting for period 2 - PATRICIA (Partitioning And Transmuter Research Initiative in a Collaborative Innovation Action)
Reporting period: 2022-03-01 to 2023-08-31
PATRICIA contributes to closing the nuclear fuel cycle by advanced recycling of spent fuel and waste burning in a dedicated accelerator driven system (ADS). In this way, together with multi-recycling of Pu for which technology exists, we can decrease the waste burden, optimize disposal and improve public acceptance paving the way toward a sustainable low carbon energy source.
Society is facing two critical problems: climate change and sustainable economic growth. Achieving climate-compatible growth requires structural reforms that support low-emission technologies including sustainalble nuclear energy. It is imperative however, that for the latter the issue of high-level waste needs to be addressed. Partitioning &Transmutation offers an attractive solution for the minimization of this waste. One of the major building blocks for the realization of P&T is the demonstration of a dedicated transmuter. MYRRHA responds to this challenge and PATRICIA supports the development of MYRRHA.
The objectives of PATRICIA are segmented in four technical domains (DMs): work on partitioning, transmutation and on MYRRHA as the transmuter demo in Europe,
Finally, the 5th domain, dissemination & communication transfers and exchanges knowledge to the project stakeholders and trains young scientists and students.
Society is facing two critical problems: climate change and sustainable economic growth. Achieving climate-compatible growth requires structural reforms that support low-emission technologies including sustainalble nuclear energy. It is imperative however, that for the latter the issue of high-level waste needs to be addressed. Partitioning &Transmutation offers an attractive solution for the minimization of this waste. One of the major building blocks for the realization of P&T is the demonstration of a dedicated transmuter. MYRRHA responds to this challenge and PATRICIA supports the development of MYRRHA.
The objectives of PATRICIA are segmented in four technical domains (DMs): work on partitioning, transmutation and on MYRRHA as the transmuter demo in Europe,
Finally, the 5th domain, dissemination & communication transfers and exchanges knowledge to the project stakeholders and trains young scientists and students.
DM1, developed a new technology to separate americium (Am) from high-level liquid waste using chemicals made up carbon, hydrogen, oxygen, and nitrogen atoms to reduce secondary waste. We improved the AmSEL process with an optimized flow sheet, designed with the PAREX+ simulation code and tested it in a laboratory setting using a 16-stage centrifugal separator. The tests show that curium (Cm) and lanthanides are well retained in the solvent, while americium was split between the solvent and the Am product, achieving a ratio of Am/Cm of about 16. Yet, 16 stages were insufficient for high Am recovery and Am/Cm ratio, necessitating further testing with a 32-stage separator. Moreover, we tested ways to make usable fuel from recycled materials, including internal gelation and co-precipitation. Porous uranium oxide microspheres were infiltrated with neodymium to form U1-yNdyO2-x beads and steps were taken to develop UAmO2 material.
DM2 assessed minor actinide-bearing fuels and improved fuel performance codes (FPCs). We did experimental and theoretical work on properties of typical fuels, such as melting temperatures, heat capacity, density, thermal expansion, and vapour pressures. Destructive and non-destructive tests of irradiated fuels were done. The data, along with thermal-mechanical properties and inert gas behaviour results, were used to enhance the FPCs TRANSURANUS and GERMINAL, and to develop the SCIANTIX meso-scale code for gas modelling. The codes were used to model Am-bearing oxide fuels under neutron irradiation, to validate the extended versions of the FPCs against past irradiation experiments. The codes, together with neutronics and thermal-hydraulics simulation tools, were used to interconnect fuel pin thermal-mechanics, coolant thermal-hydraulics, and neutronics feedback to do a multi-physics design of MYRRHA transient irradiation scenarios. We also did 4 studies on Am-bearing fuels in normal and off-normal conditions, such as operating a transmutation target in MYRRHA with (U,Pu,Am)O2 containing 0.49% to 5% Am and, behaviour during a beam jump transient when the proton beam switches on. Preliminary designs for future transient tests in the HFR and BR2 were made. An evaluation of facilities for transient testing with Am-bearing fuels aims to identify those capable of reproducing fast transients not achievable by Material Testing Reactors.
DM3 focuses on MYRRHA core safety. An integral test, exposing a 7-pin wire-spaced, heated bundle, representative of MYRRHA, to lead-bismuth eutectic (LBE) for nearly 2000 hours under nominal conditions was done in the MEXICO loop. The data validated coupled thermal hydraulics and chemistry models. We found no plugging risk from corrosion products and the onset of flow reduction was only seen above the precipitation limits of PbO. MYRRHA cladding materials were corroded in static LBE, with samples used to compare mechanical strength loss with cladding thickness. First results show no further mechanical strength loss beyond those caused by the loss of the corroded layer.
Detailed analyses of MYRRHA representative UO2 segments used in power transient experiments revealed small deformations by PCMI. We will cut some pins in segments to investigate their internals. Finally, a heated rod bundle with a defined blockage was made and investigated experimentally to study thermal hydraulics of a fuel bundle under extreme conditions, accompanied by numerical modelling to transfer the results to the conditions of a real reactor. A bend rod bundle experiment is in progress.
DM4 addresses the safety of the MYRRHA Accelerator-Driven System. Data from comparable accelerators showed that for MYRRHA, reliability should be 100 fold better thus stressing the need to test reliability and fault tolerance schemes on the MINERVA accelerator. The acceptable accelerator downtime of 3 seconds should be re-evaluated. Studies on beam window breakage and radionuclide release identified signals corresponding to evaporated mercury, requiring further analysis to understand the observed mechanisms of evaporation and transport. Construction, licensing, and testing of the POVACS facility, to investigate polonium's sticking on stainless steel, are nearing completion.
WP11 studies thermal hydraulics of the reactor pool. We tested the transition from forced to-natural-circulation in CIRCE and ESCAPE (completed), and developed numerical models in parallel. A heat transfer model was implemented and validated against DNS data, with a wall-function for turbulent heat flux developed based on the Jayatilleke model, implemented in OpenFOAM v8.
WP12 did experiments and calculations to study the chemical processes in the coolant to create and validate models of chemical reactions in LBE. We studied radioactivity release, essential for MYRRHA's safety assessment, progressed in understanding the release behaviour of polonium, got experimental and theoretical results for the release of tellurium from LBE and the stability of associated molecules. Moreover, we prepared LBE mixtures with ruthenium and studied their evaporation, as well as the evaporation of thallium from LBE in the presence of iodine.
DM5 deals with project management and dissemination. Many publications and presentations were issued, several PhD students were hired and a heavy metal summer school for education and networking was set up.
DM2 assessed minor actinide-bearing fuels and improved fuel performance codes (FPCs). We did experimental and theoretical work on properties of typical fuels, such as melting temperatures, heat capacity, density, thermal expansion, and vapour pressures. Destructive and non-destructive tests of irradiated fuels were done. The data, along with thermal-mechanical properties and inert gas behaviour results, were used to enhance the FPCs TRANSURANUS and GERMINAL, and to develop the SCIANTIX meso-scale code for gas modelling. The codes were used to model Am-bearing oxide fuels under neutron irradiation, to validate the extended versions of the FPCs against past irradiation experiments. The codes, together with neutronics and thermal-hydraulics simulation tools, were used to interconnect fuel pin thermal-mechanics, coolant thermal-hydraulics, and neutronics feedback to do a multi-physics design of MYRRHA transient irradiation scenarios. We also did 4 studies on Am-bearing fuels in normal and off-normal conditions, such as operating a transmutation target in MYRRHA with (U,Pu,Am)O2 containing 0.49% to 5% Am and, behaviour during a beam jump transient when the proton beam switches on. Preliminary designs for future transient tests in the HFR and BR2 were made. An evaluation of facilities for transient testing with Am-bearing fuels aims to identify those capable of reproducing fast transients not achievable by Material Testing Reactors.
DM3 focuses on MYRRHA core safety. An integral test, exposing a 7-pin wire-spaced, heated bundle, representative of MYRRHA, to lead-bismuth eutectic (LBE) for nearly 2000 hours under nominal conditions was done in the MEXICO loop. The data validated coupled thermal hydraulics and chemistry models. We found no plugging risk from corrosion products and the onset of flow reduction was only seen above the precipitation limits of PbO. MYRRHA cladding materials were corroded in static LBE, with samples used to compare mechanical strength loss with cladding thickness. First results show no further mechanical strength loss beyond those caused by the loss of the corroded layer.
Detailed analyses of MYRRHA representative UO2 segments used in power transient experiments revealed small deformations by PCMI. We will cut some pins in segments to investigate their internals. Finally, a heated rod bundle with a defined blockage was made and investigated experimentally to study thermal hydraulics of a fuel bundle under extreme conditions, accompanied by numerical modelling to transfer the results to the conditions of a real reactor. A bend rod bundle experiment is in progress.
DM4 addresses the safety of the MYRRHA Accelerator-Driven System. Data from comparable accelerators showed that for MYRRHA, reliability should be 100 fold better thus stressing the need to test reliability and fault tolerance schemes on the MINERVA accelerator. The acceptable accelerator downtime of 3 seconds should be re-evaluated. Studies on beam window breakage and radionuclide release identified signals corresponding to evaporated mercury, requiring further analysis to understand the observed mechanisms of evaporation and transport. Construction, licensing, and testing of the POVACS facility, to investigate polonium's sticking on stainless steel, are nearing completion.
WP11 studies thermal hydraulics of the reactor pool. We tested the transition from forced to-natural-circulation in CIRCE and ESCAPE (completed), and developed numerical models in parallel. A heat transfer model was implemented and validated against DNS data, with a wall-function for turbulent heat flux developed based on the Jayatilleke model, implemented in OpenFOAM v8.
WP12 did experiments and calculations to study the chemical processes in the coolant to create and validate models of chemical reactions in LBE. We studied radioactivity release, essential for MYRRHA's safety assessment, progressed in understanding the release behaviour of polonium, got experimental and theoretical results for the release of tellurium from LBE and the stability of associated molecules. Moreover, we prepared LBE mixtures with ruthenium and studied their evaporation, as well as the evaporation of thallium from LBE in the presence of iodine.
DM5 deals with project management and dissemination. Many publications and presentations were issued, several PhD students were hired and a heavy metal summer school for education and networking was set up.
PATRICIA ontributes to a solution for used fuel by recycling and fuel reuse. Untreated used fuel needs geological storage for several hundred thousand years. Societal acceptance can be improved and constraints are easier to achieve if the storage time can be significantly cut back and the storage volume reduced, independent of the technical status of the depository. Overall, the project made significant progress in developing technologies and processes for the efficient and safe management of minor actinides in nuclear reactors, with a particular focus on MYRRHA. The research spanned from chemical separation and fuel fabrication to safety assessments and performance modelling, laying the groundwork for future advancements.