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"Methodology, Analysis and eXperiments for the ""Safety In MYRRHA Assessment"""

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

Executive Summary:
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 goal of MAXSIMA is to contribute to the "safety in MYRRHA" assessment.

The MAXSIMA project consists of five technical work-packages. The first contains safety analyses to support licensing of MYRRHA. Design-based, design extended and severe accident events have been studied with a focus on transients potentially leading to fuel pin failures. Fuel assembly blockage and control system failure are the least unlikely events leading to core damage. A thermal-hydraulic study of different blockage scenarios of the fuel bundle as well as tests of the hydrodynamic behavior of a new buoyancy driven control/safety system have been carried out. Both were supported by numerical simulations. Safety of the Steam Generator (SG) is treated by looking at consequences and damage propagation of a SG Tube Rupture event (SGTR) and by characterizing leak rates and bubble sizes from typical cracks in a SGTR. Additionally, a leak detection system and the drag on bubbles travelling through liquid LBE was studied. Transient test methodology for MYRRHA type fuel and experimental devices have been implemented at the ACPR-TRIGA reactor in Romania. The goal of the tests is to establish the failure threshold of fast reactor fuel pins during fast power transient. MAXSIMA include validation experiments for safety computer codes involving core damage scenarios with high temperature MOX-LBE interactions. Fuel-coolant-clad chemistry is studied up to 1700°C and a core melt experiment in a reactor is prepared to assess the interaction of LBE with molten fuel. Following the Fukushima return of experience, effort was put on development of enhanced passive safety systems for decay heat removal and on confinement analyses for HLM systems. A separate work package is dedicated to education and training. Beside workshops, lecture series and training sessions, a virtual-safety simulator software was developed.
Project Context and Objectives:
The aim of WP2 ‘Safety Analysis in support of MYRRHA‘ activities consists in the study of a series of transients identified for the MYRRHA reactor (a new nuclear facility for material testing and isotope production to be built in SCK•CEN domain) in its updated design configuration and in providing state-of-the-art safety analyses that can serve as input to the Belgian Safety Authority looking forward to the facility licensing activities.
The main target consists in covering the MYRRHA safety analysis from different perspectives:
• Design of the MYRRHA core (and required shielding studies) using 3-D methods
• Study of a complete list of accidental events and analysis of input data uncertainty propagation in the safety-relevant output parameters
• Analysis of a number of severe accident scenarios potentially leading to core disruption.

Each main topic listed above has been studied in a specific Task within the Work Package.
The activities of the three Tasks are deeply intertwined. The data exchange between the Tasks is required and several activities are developed across different tasks: the comparison of different approaches towards the solution of a specific problem provides an interesting insight from different perspectives.
Nevertheless, it is possible to make a broad correspondence between every Task and the specific topic analyzed. In detail, the WP2 activities are organized in three Tasks:
• Task 2.1: Neutronic and Shielding Analysis in Support of Safety Studies
• Task 2.2: Transient Analyses Using System Codes
• Task 2.3: Severe Accident Analysis
Moreover, collaborations with other Work Packages have been foreseen, in particular with WP3 (sharing the safety limits adopted for core damage), WP4 (providing an updated design configuration for experimental facilities setup) and WP7 (providing first list of requirements for an operator training simulator).

WP3 ‘Core Component Safety' is dedicated to give a detailed insight into the safety aspects of the fuel assemblies and the control and safety rods of the reactor core. In the fuel assembly, the cooling of a partially blocked fuel rod bundle was experimentally investigated. A second experiment was carried out to validate the correct movement of buoyancy driven control and safety rods. Both experiments were numerical supported by CFD simulations.

WP4 ‘Steam Generator and cooling safety’: One of the most challenging issue related to the technological demonstration of a Lead-cooled Fast Reactor or an Accelerator Driven System (cooled by heavy liquid metal), concerns the possibility to install the primary heat exchanger (or steam generator) directly into the primary pool, avoiding any intermediate circuit as done in the past for sodium-cooled fast reactor.
According to that, in frame of the MAXSIMA Project, and in particular as goal of the WP4, it has been proposed to demonstrate the level of safety of installing a steam generator into the primary pool, considering as basic scenario the rupture of a pressurized tube.
In particular, a large scale experiment, in CIRCE pool, (the largest heavy liquid metal pool operating in Europe) has been designed, implemented and performed aiming at characterizing the Steam Generator Tube Rupture (SGTR) event in a configuration relevant for MYRRHA.
In parallel numerical tools have been verified and validated to support the design phase as well as the safety assessment of such solutions.
Moreover, to study the detectability of a tube leakage (leak before break) to prevent large breaks, a devoted experimental campaign has been performed on a separate effect facility (LIFUS5) aiming at characterize, as function of the steam bubbles size released by a typical crack, the bubble velocity and migration in a liquid metal pool, and the capability of the developed instrumentation to detect such occurrence.
Also in this case numerical tools have been adopted to demonstrate their capability to predict and simulation the related phenomena.
Finally, in TALL facility, the bubble migration has been investigated providing experimental data on drag coefficient of gas bubbles travelling through liquid metal, needed to improve numerical models aiming at simulating such phenomena.

WP5 ‘Fuel safety’: The objectives of WP5 include test methodology development and validation for transient testing of MYRRHA type fuel for the determination of the pin failure threshold, investigation of fuel-coolant and fuel-coolant-clad interactions and development of a conceptual design for an experiment related to severe accident analysis of MYRRHA type fuel in the IGR reactor at NNC (Kazakhstan).
The TRIGA Annular Core Pulsing Reactor (ACPR) operated by RATEN ICN (Romania) is proposed as testing facility for the fast reactor fuel safety experiments. The methodology and experimental set-up validation is based on irradiation tests on uranium dioxide test fuel segments designed specifically to allow for the simulation of MYRRHA MOX fuel that has attained burn-up. The design of the pins is such that the conditions of the fuel during the experiment are close to the limits for PCMI.
For the assessment of severe nuclear accidents involving a fuel clad failure it is important to address a possible interaction between the MOX fuel and the coolant. Fuel-coolant-clad chemistry is studied up to 1700°C. Temperature points before and above the melting temperature of the clad material are targeted.
The requirements for the experiment in the IGR reactor are based on key licensing issues for MYRRHA fuel (early phase core degradation). Such experiment in reactor is prepared to assess the interaction of LBE with molten fuel.

The main objectives of the WP6 ‘Enhanced Safety by Design for HLM reactors’ was to develop an enhanced innovative passive safety system for Decay Heat Removal (DHR) on the basis of a critical review of the conceptual design of safety systems already developed in the past in the frame of the development of heavy liquid metal cooled reactors. For such reactors (with a high or relatively high melting point) the systems dedicated to heat removal should also guarantee that the primary coolant is not brought to the so-called freezing (solidification) condition. The work developed through the analysis of the innovative passive safety system proposed by Ansaldo to prevent coolant freezing. Simulations have been carried out by state of the art computational tools (RELAP5 and TRACE) showing that the system is able to fulfil the expectations.
As an additional goal the Work Package included the evaluation of source terms for MYRRHA as well as a the containment analysis using the most advanced tools available, for the most challenging transients that were identified. The simulation models for the containment analyses have been verified by means of a comparison between different codes (CONTAIN and MELCOR) on the basis of mass and energy releases produced by system codes.
Dedicated source term for MYRRHA (within the present level of knowledge) have been developed and simulations of containment behavior for the selected transients shown consistent results between CONTAIN and MELCOR models.

WP7 ‘Education and Training’: Heavy liquid metal (HLM) cooled systems, such as MYRRHA, ALFRED and ELECTRA have several advantageous safety features, such as potential for decay heat removal by natural convection, benign coolant-water interaction and fission product retention. Among the drawbacks are coolant activation, the need for a strict coolant chemistry control and in the case of pure lead, the potential for coolant freezing.
One objective of this WP was to train young scientists and students, aiming to give them functional skills and insights pertaining to safe operation of HLM cooled reactors. For this purpose, a virtual simulator environment was developed. Students could train management of incidents and accidents in HLM cooled systems. Prior to the simulator training the students participated in seminars where lectures on the safety of heavy liquid metal systems formed the basis for student projects on the topic.
This WP was also responsible for dissemination of knowledge obtained in the project. Two workshops were organized next to the preparation of materials for the public part of the project website.

Project Results:
The main MAXSIMA topics and obtained results can be divided in five categories, corresponding to the technical Work Packages. Each category refers to the relevant deliverables containing the detailed data, experiment or analysis.
Safety analysis in support of MYRRHA
• Neutronic and shielding analysis in support of safety studies (D2.1 D2.2)
• Transient analyses using system codes (D2.3 D2.4)
• Severe accident analyses (D2.5 D2.6)

Core component safety
• Thermal hydraulic fuel assembly blockage experiments (D3.1 D3.2 D3.3 D3.4)
• Safety Rods system tests in HLM (D3.5 D3.6)
• Fuel Blockage simulation (D3.7)
• CFD simulation of safety/control rods system (D3.8 D3.9)

Steam generator and cooling safety
• SGTR Propagation (D4.1 D4.2 D4.3)
• SGTR Bubble Characteristics (D4.4 D4.5 D4.6)
• Bubble transport validation (D4.7 D4.8)

Fuel safety
• Transient testing of MYRRHA fuel (D5.1 D5.2 D5.3 D5.4 D5.5 D5.6 D5.7 D5.8 D5.9 D5.10 D5.11 D5.12 D5.13)
• Fuel-coolant interaction (D5.14 D5.15 D5.16)
• Conceptual design study for experiment in the IGR reactor (D5.17)

Enhanced Safety by Design for HLM reactors
• Safety systems (D6.1)
• Development of innovative passive safety systems (D6.2)
• MYRRHA containment analysis (D6.3)

A detailed overview of the main scientific and technical results is available in the PDF attached (MAXSIMA final report_main S-T results).

Potential Impact:
Every Work Package has a clearly defined outcome with a direct impact on the safety assessment of heavy metal cooled reactors like MYRRHA. The detailed neutronic characterization provides 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, was assessed. Also, the safety studies have been 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 have been 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 represents an important international database for CFD computer code validation.

The effects of a SGTR event in terms of mechanical damage to in-vessel structures have been assessed experimentally. In addition, input information for the evaluation of the migration of steam bubbles in the melt by numerical simulations was be generated. The leak rate and bubble sizes from typical cracks have been characterised and the friction coefficient of gas bubbles moving in liquid LBE determined. Moreover, an early leak detection system was 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 was 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 clarifies 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 were 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 was 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.

In the frame of WP 7, dissemination of knowledge obtained in the project took place:
• The first MAXSIMA workshop was successfully implemented in Karlsruhe on October 7-10, 2014 in Karlsruhe, including 5 keynote lectures and 41 technical presentations. With 85 registered participants, the workshop succeeded to enhance the information exchange within the international community working on heavy-liquid-metal systems.
• The second MAXSIMA workshop was organized by KTH in Jukkasjärvi (The Icehotel) in Sweden on February 23-26, 2016. In total, nineteen presentations were displayed.
• The MAXSIMA Lecture Series “Lecture notes on safety of HLM systems” was held at the Karlsruhe Institute of Technology from March 24-27, 2015. The main goal was to provide comprehensive background knowledge on safety of HLM systems. Also safety aspects of light water reactors was covered. In total, fourteen lecture notes were presented.
During the second 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, was presented. Within this virtual simulator environment, students can train management of incidents and accidents in HLM cooled systems.

The records of the publications and dissemination activities are as follows:

publications in journals: 10 + 1 being reviewed
organization dedicated website: 1
organization workshops and Lecture Serie: 3
contributions (presentations/posters) in international conferences/workshops: 23
contribution (presentation) in a national workshop (Germany): 1
contributions to Summer Schools: 4
contributions (presentations/posters) in the MAXSIMA workshops: 42
contributions in Lecture Series:16

List of Websites:
http://maxsima.sckcen.be/