Community Research and Development Information Service - CORDIS

Periodic Report Summary 3 - MAXSIMA (Methodology, Analysis and eXperiments for the "Safety In MYRRHA Assessment")

Project Context and Objectives:
The MAXSIMA project's objective is to provide a contribution to the safety in MYRRHA assessment. MYRRHA (Multi purpose hYbrid Research Reactor for High tech Applications) is a flexible fast spectrum research reactor. It is a flexible fast spectrum research reactor (50-100 MWth), conceived as an accelerator driven system (ADS), able to operate in sub-critical and critical modes. It contains a proton accelerator of 600 MeV, a spallation target and a multiplying core with MOX fuel, cooled by liquid lead-bismuth (Pb-Bi).

The Strategic Research Agenda of the EU Sustainable Nuclear Energy Technical platform requires new large infrastructures for its successful deployment. MYRRHA has been identified as a long term supporting research facility for all ESNII systems and as such put in the high-priority list of ESFRI.

The consortium of participants in the project included 13 institutions from all over Europe: ENEA (Italy); KIT (Germany); ANSALDO (Italy); GRS (Germany); NRG (Netherlands); CRS4 (Italy); KTH (Sweden); HZDR (Germany); RATEN (Romania); CHALMERS (Sweden); NNC (Kazakhstan); CIEMAT (Spain) and SCK•CEN (Belgium).

MAXSIMA consists of 7 work packages. Work Package 1 is dedicated to the general coordination of the consortium, in particular the preparation and organization of the project management meetings, the monitoring delivery dates of deliverables and milestones. The status of all the Work Packages and tasks has been monitored by means of the PCC and GB meetings.

The main objectives of Work Package 2 is to execute a global safety analysis in support of the licensing process, for this reporting period a list of protected and unprotected transients under investigation, a working range of the MYRRHA core and basic neutronic parameters for safety analyses and a shielding and activation study of MYRRHA in both critical and sub-critical mode.

Work Package 3 is dedicated to the experimental validation of the control/safety rods and the experimental validation of design.

The activities in Work Package 4 are focused on the experimental validation of safety codes for steam generator tube rupture propagation and associated steam bubble transport.

The objectives of Work Package 5 include the set-up of the design requirements for the fuel pin and the rig to be tested during an irradiation campaign in the TRIGA reactor. For the assessment of severe accidents involving a fuel clad failure in MYRRHA, possible interaction between the MOX fuel and the coolant needs need to be studied experimentally.

Work Package 6 is dedicated to the enhanced safety by design for HLM reactors. The main objective for the first reporting period is the identification of enhanced innovative passive safety systems. The start is a critical review of the safety systems conceptual design already developed, where the main advantages and drawbacks will be highlighted and a number of desirable features identified. This review is the basis for a first conceptual design of the systems, discussed and agreed by the partners.

Work Package 7 is devoted to Education and Training. The lecture series on safety of heavy liquid metal – originally planned for the first reporting period – has been postponed to 2015. A plan for dissemination was proposed by KTH. The development of a neutronics engine for a virtual safety simulator software is also included in the objectives for this period.
Project Results:
A global safety analysis in support of the MYRRHA licensing process has been carried out. During this reporting period the final reports of the transient analysis and severe accident analysis in critical and sub-critical mode have been released.

The coolability of a fuel bundle in accident scenarios by experimentally studying the effects of channel blockage has been realised. A second experiment aiming to validate a buoyancy-driven control rod system was carried out. This system has a safety function in emergency scenarios in the sense that it will act as a diverse backup of the nominal safety rods. Both activities were supported by numerical simulations.

Experimental validation of safety codes for steam generator tube rupture (SGTR) propagation and associated steam bubble transport have been carried out. Essential aspects of a SGTR were studied. A propagation experiment was done to investigate damage to structures in the reactor and possible damage propagation. Leak rates and bubble sizes coming from typical tube cracks have been determined and an early detection system was tested. The drag coefficient of gas bubbles moving through the liquid metal has been measured.

A first irradiation campaign took place in a TRIGA reactor. For the assessment of severe accidents involving a fuel clad failure in MYRRHA, possible interaction between the MOX fuel and the coolant were studied experimentally. The conceptual design study for experiments with MYRRHA fuel in the IGR reactor from NNC has been completed. The objective of these experiments is to gather representative data of key-licensing phenomena for the MYRRHA fuel.

In order to enhance the safety by design for HLM reactors, a conceptual design for an innovative DHR system to be used in any HLM-cooled pool-type reactor was developed and the analysis of the innovative passive safety system to prevent coolant freezing was performed using computer code. During this reporting period, an evaluation of the containment response for MYRRHA in terms of pressure and temperature to a mass and energy release from the side of the secondary cooling was carried out.

The first MAXSIMA workshop was successfully implemented in Karlsruhe on October 7-10, 2014 in Karlsruhe, including 5 keynote lectures and 41 technical. The second MAXSIMA workshop was successfully implemented by KTH on February 23-26, 2016 in Jukkasjärvi, Sweden. During this workshop, a first beta version of an application for mobile platforms, the so-called ‘MAXSIMA Simulator’, which simulates the transient behaviour of Generation-IV lead-cooled fast reactors for safety-informed pre-design and didactical purposes, has been presented.
Potential Impact:
Every Work Package has a clearly defined outcome with a direct impact on the safety assessment of MYRRHA. The detailed neutronic characterization will supply an important envelope for safety studies. The impact of some accident situations on the neutronic features of the core like steam or water ingress into the core, core compaction, control rod ejection, beam window breakage, will be assessed. Also, the safety studies will be extended to an evaluation of the radiological safety and protection. Finally, the core reactivity and related reactor power variations due to all major reactivity feedback effects, including relocation of core materials under accident conditions will be assessed.

The work defined on core component safety is focused on two main aspects of safety aspects in the reactor core. The first deals with the coolability of a fuel bundle in accident scenarios by experimentally studying the effects of channel blockage. The second experiment aims to validate a buoyancy-driven control rod system that in emergency scenarios will have a safety function. The results of these experiments will represent an important international database for CFD computer code validation.

The effects of a SGTR event in terms of mechanical damage to in-vessel structures will be assessed experimentally. In addition, input information for the evaluation of the migration of steam bubbles in the melt by numerical simulations will be generated. The leak rate and bubble sizes from typical cracks will be characterised and the friction coefficient of gas bubbles moving in liquid LBE will be determined. Moreover, an early leak detection system will be qualified. Early detection of a leak can prevent further propagation that could eventually lead to a full rupture of the tube. These activities will contribute to support the design of heavy liquid metal fast reactor, with main focus on MYRRHA nuclear system and its pre-licensing.

Although the new designs like MYRRHA, benefit from technology and experience of past sodium fast reactors, the modification of the coolant type to LBE and the modified operating conditions necessitates the (re)evaluation of safety related parameters of reactor core and fuel. The difference in coolant type will have direct consequences for the coolant-cladding interaction and fuel-coolant interaction during severe accident conditions. Works focused on transient irradiation experiments by delivering a power pulse to a fuel segment in a dedicated capsule mounted in the TRIGA reactor as well as on the pre-engineering of a core melt experiment in LBE for MYRRHA type fuel are planned. Experimental verification will be made of possible fuel coolant chemical interactions which could occur in case of a cladding breach. The complementary results obtained in MAXSIMA and the outcomes of other projects like SEARCH will clarify this issue for LBE and Pb cooled fast reactor systems and have a large impact on further safety assessment studies not only for MYRRHA but also for ALFRED, LFR or other industrial scale ADS.

Passive safety systems will be proposed to reach first a conceptual design of the systems and proceed to a more detailed and verified design to be tested for the different main references of the EU LFR roadmap, i.e. the MYRRHA, ALFRED and ELFR. A detailed evaluation of source terms as well as a containment analysis for the reference transients that will be identified, is also an outcome of the present project.

With the growing scarcity of fossil fuels and the problem of CO2 emissions, energy is one of the critical challenges society faces. The heavy liquid metal cooled Gen IV systems show promising features to solve this issue by using fuel more efficiently while producing less waste. By investigating the safe behaviour of the fuel and the coolant in MYRRHA in its role as the European Technology Pilot Plant (ETPP) for the heavy liquid metal cooled reactor, the MAXSIMA project directly addresses these issues.
List of Websites:
http://maxsima.sckcen.be/

Reported by

STUDIECENTRUM VOOR KERNENERGIE
Belgium

Subjects

Nuclear Fission
Follow us on: RSS Facebook Twitter YouTube Managed by the EU Publications Office Top