1. ESNII+ summer school, KTH, Stockholm, May 20, 2014
Swedish contributions to the safety of ASTRID reactor: multi-grant project at KTH
Diana Caraghiaur1, Augusto Hernandez Solis1, Ranjan Kumar1, Sebastian Raub1, Pavel Kudinov1, Weimin Ma, Nathalie Marie2, Christophe Journeau2, Laurent Trotignon2, Frédéric Bertrand2 and Sevostian Bechta1
1 - Division of Nuclear Power Safety at Royal Institute of Technology (KTH)
2 - French Alternative Energies and Atomic Energy Commission - Commissariat à l'énergie atomique et aux énergies alternatives (CEA), Cadarache
ESNII+ summer school, KTH, Stockholm, May 20, 2014

2. VR framework grants to support first long term direct collaboration between France and Sweden in nuclear energy
The purpose is to strengthen nuclear research and to stimulate the development of Jules Horowitz Research Reactor (JHR) and sodium-cooled prototype reactor (ASTRID):
Multi-grant projects: 60 MKr, in two calls with deadlines in and 2013
PhD students/Postdocs: - Work at CEA, Cadarache - Enrolled in Sweden and do a Swedish PhD/Postdoctoral project - Both Swedish and French supervisors

3. VR Multi-Project Grants in Nuclear Energy Research
3 multi-grant projects funded by the Swedish Research Council in the spring of 2012 (projects in collaboration with CEA, France – French Alternative Energies and Atomic Energy Commission):
DEMO-JHR (coordinator: Prof. Christophe Demazière, Chalmers): 3 PhD projects including 1 at KTH: DEPTHS, Development of Procedures of Thermal-Hydraulic Simulations for JHR
ASTRID core physics and diagnostics (coordinator; Prof. Imre Pázsit, Chalmers): 4 PhD projects including 1 at KTH: ALDESA, Acoustic Leak DEtection in Sodium Applications
ASTRID safety (coordinator: Prof. Sevostian Bechta, KTH): 1 PhD + 3 post-doc projects.

4. VR Multi-Project Grants in Nuclear Energy Research (2)
2nd VR call project of 2014:
Dr Staffan Jacobsson Svärd, Uppsala universitet: Assessing fuel behavior in the sodium-cooled fast reactor ASTRID, 2.5 MSEK
Prof. Janne Walenius, KTH: two PhD projects on Thermodynamic assessments of relevance for fuel- cladding interaction and on Modelling of fission product transport in MOX fuel, 6 MSEK
Prof. Imre Pazsit, Chalmers: Neutronic modeling of control rod withdrawal, 2.5 MSEK

5. ASTRID-Safety multi-grant project at KTH
Aimed at safety improvement of ASTRID SFB including severe accident prevention and mitigation
WP1 – Corium retention
1 PhD student, 4 years
WP2 – Simulation of an early phase of a severe accident 1 post-doc, 2 years
WP3 – Probabilistic safety analysis
1 post-doc, 2 years
WP4 – Analysis of severe accident scenarios with simplified models

6. WP1: Corium retention PhD student: Sebastian Raub
Supervisors: Pavel Kudinov (KTH)
and Christophe Journeau (CEA)
WP is aimed at integration of CEA and KTH experience in LWR and SFB safety and improvement of severe accident management (SAM) of ASTRID reactor

7. Research tasks
Contribution into ASTRID core catcher with the studies of:
- melt fragmentation in sodium coolant and corium debris bed formation,
- long term coolability of corium debris by sodium natural convection
Application of models and tools developed in LWR safety for analysis of to usage for Sodium
The current approach is building on KTH-Expertise in Severe Accidents in a typical Swedish Light Water Reactor.
The mathematical underpinning of existing Software Tools, developed at KTH is analyzed for assumptions and deductive steps are no longer valid or need adjustment due to the shift from water to liquid sodium and their respective sets of physical properties.
Incorporate the adjustments into the existing code structure
Evaluate modified Code
Find under which range of parameters and boundary conditions the code will perform according to expectations ( debris mass flow into coolant pool, debris temperature, etc)
Add Improvements to existing Code Package
Possibilities include enhanced heat transfer and vapor production capabilities, as well as debris heap settling and

8. First ideas about modeling
3 Major Modules:
Flow in Porous Media/Evaporation:
Solves Ergun’s Filtration equation for flow in a packed bed
Solves Phase continuity with flow
Does not attempt heat transfer calculations  Heat Release goes completely in evaporation
Flow in the coolant pool:
Solves Equation of continuity and
Turbulent Equation of Momentum, using the
k – ε turbulence model
Melt Particle Motion and Depositon:
Lagrangian model for each particle
Particles do not interact
Velocity vector of particles consists of 2 parts
Velocity field for liquid phase with corrections for buoyancy force
Random vector with Gaussian probability distribution, magnitude of variance tied to turbulence
Provides corrections to the flow fields due to:
pressure differentials
density changes
Phase changes
Provides flow fields for both phases
Provides flow fields for both phases
Provides drag force on particles
Debris Bed From by particle depositon

9. WP2: Simulation of an early phase of a severe accident
Postdoctoral researcher: Augusto Hernandez-Soliz
Supervisors:
Weimin Ma (KTH)
Laurent Trotignon,
Pierre Gubernatis (CEA)
WP is focused on further development of SIMMER-III code for the initial phase of SA with core melting and relocation

10. Motivation
The core design of ASTRID had posed many challenges for the modeling of neutronic and TH phenomena
| PAGE 10

11. ASTRID fuel pins Unprotected Transient of Power (UTOP)
We are focused on simulating what happens at the fuel pin at the begining of an accident
State-of-the art safety analysis relies on computer codes (e.g. SAS-SFR, SIMMER) in order to understand what happens during an unexpected reactor transient
8,5 mm
Fissile zone of fuel rods :
0,4 kg of fuel + 0,1 kg of steel
Relatively high internal pressure (100b)
Fuel assambly(271 rods) :
95 kg of fuel + 53 kg of steel
SNa  1/3 Sth

12. Fuel behaviour during a TOP
The power transient (initial or induced) causes a rapid heating of the fuel, creating a cavity inside the pin. Degradation is driven by the melting of UO2, and due to the dilatation and rapid pressurization of the cavity up to mechanical failure
On such a sequence, we look to represent the so called early stage of the primary phase of the transient, i.e.:
At the pin level: Fuel heating and meelting, evolution and pressure of the cavity up to cladding rupture and ejection of molten fuel to Na
Transient of Power
| PAGE 12

13. Initiating (primary) stage
Solution scheme
Transition
(1° rupture of can wall)
Initiating event
Secondary stage
Initiating (primary) stage
Transition stage
Irradiation
Current scheme
SAS one-pin h = point-kinetic
SIMMER III or IV : 1 or 2 mesh per assembly
Point kinetics
Limitations on handling CFV core geometry
Cannot handle multi-pin modeling per channel,
(i.e. one average pin/channel)
Drawbacks
Neutronics = PARIS
Point-kinetic OR spatial
SIMMER III or IV : 1 or 2 mesh per assembly
SIMMER-III*
Multi-classes in // and Multi-pins
+ DPIN
GERMINAL
Cathare : TH of the reactor loop
New scheme
| PAGE 13

14. SIMMER-III code
The code that will be used in this project corresponds to the SIMMER-III (V. 3E) code
Developed by KIT (Germany), JNC (Japan) and CEA Cadarache to study the consequence of core disruptive accidents in SFRs
| PAGE 14

15. Detailed PIN-I model of the SIMMER-III code
The best possible way to model a fuel pin can be found in DPIN-1:
Can handle annular pellets
A mesh of 11 nodes can be defined for the pellet in order to model accurately temperature profiles
DPIN-1 (<11 grids)
Pcav
Grid (<11)
Cp(Ti), ri, li,
Porosities ei
Local concentrations of FPs
Cavity = Pcav (PU02+PPF), Tcav.
Thermal-hydraulics of UO2 are not
modeled in the cavity
Simplified mechanical model of the fuel
| PAGE 15

16. Objectives of the project
Attempts to model the motion of the molten fuel inside the cavity have been carried out in DPIN-2
Altough not very succesfully
Therefore, the main idea of the project is to improve the DPIN-1 model of SIMMER-III
By trying to implement a time-dependent and axial in-fuel model within the cavity, where the thermal-hydraulics effects are taken into account
Retro-engineering can be performed based on the in-fuel
motion models of other codes (such as SAS-SFR)
cavity
| PAGE 16

17. Sensitivity analysis in SAS-SFR
In order to do retro-engineering from SAS-SFR into SIMMER-III, the most important model parameters should be known in advance in order to simplify the work
The aim is to rank the importance of the different input parameters towards a certain output
Statistical methods can be used for such sensitivity analysis (SA)
| PAGE 17

18. Some sensitivity analysis results obtained with SAS-SFR
TOP
Fuel conductivity (1-Sigma = 1%), Normal PDF
Maximum axial clad. temp. after 100 calculations
| PAGE 18

19. Some sensitivity analysis results obtained with SAS-SFR (2)
TOP
Porosity (1-Sigma = 1%), Normal PDF
Maximum axial clad. temp. after 100 calculations
| PAGE 19

20. Some sensitivity analysis results obtained with SAS-SFR (3)
TOP
Gap Heat Transfer Coeff. (1-Sigma = 1%), Normal PDF
Maximum axial clad. temp. after 100 calculations
MOST DOMINANT PARAMETER
| PAGE 20

21. Future plans SAS-SFR SIMMER-III
Build a general vision of the SAS-SFR approach + phenomena
In-fuel motion model
Thermo-mechanical cavity model
Understand the SIMMER-III approach of fuel pin degradation modeling
Pin Model
EJECT
DEFORM
Pin Model
SPIN
DPIN
Compute selected CABRI tests (TOP)
E5/E7
LT2/PF2
Improved DPIN-I
In-fuel motion model
| PAGE 21

22. WP3: PSA of ASTRID system design
Postdoctoral researcher: Ranjan Kumar
Supervisors:
Pavel Kudinov (KTH)
Frederic Bertrand (CEA)
WP is aimed at Dynamic PSA of ASTRID Decay Heat Removal System taking into account failed component recovery

23. Probabilistic Safety Assessment of ASTRID DHR Systems
ASTRID has a significant boiling margin in normal operation (more than 300°C) together with a high thermal inertia of the primary system (advantage over PWR).
Decay heat removal systems mainly use
air as a cold source and they are based on forced and natural convection, which
allows passive mode of systems operation.
DHR system must be practically free from the loss of the decay heat removal function.
Probabilistic Safety Assessment of ASTRID DHR Systems

24. PSA Research Objectives
To evaluate Core Damage Frenquency for current ASTRID DHR systems design and analyse potential accident scenarios with consideration of recovery of failed items.
To perform PSA of current design of ASTRID DHR systems taking into account both probabilistic and deterministic approaches.
To demonstrate and improve the design for practically failure free DHR function.
Probabilistic Safety Assessment of ASTRID DHR Systems

25. Schematic Diagram of ASTRID DHR System
ASTRID DHR systems consist of four types of DHR systems:
4 loops of S1 used in normal and accident conditions (100% for first 3 days)
2 active loops of S2 (2 x 100%)
2 passive loops of S3 (2x 100%)
2 loops in the bottom of the reactor vessel S4(2x 50%)
Probabilistic Safety Assessment of ASTRID DHR Systems

26. Grace Period for Components Recovery
Figure shows the long term sodium temperature calculations taking into account the DHR system operation.
Grace period depends on the sequences of failure and their effects on the sodium temperature rise.
S2/S3 S4
Probabilistic Safety Assessment of ASTRID DHR Systems

27. CEA-KTH aproach: A PSA Level-1 Method with Repairable Components
A PSA Level-1 Method with Repairable Components is proposed and implemented by exploiting the grace period using partial dynamic Event Tree and Fault Tree (To be presented in ESREL2014 conference)
C1-C5 are either OK or NOT OK based on users defined decoupling criteria
Probabilistic Safety Assessment of ASTRID DHR Systems

28. A Repairable Fault Tree in Proposed PSA level I
(with recovery)
(without recovery)
𝜆 2
𝜆 1 , 𝜇 1
𝜆 4
Top Event
𝜆 5 , 𝜇 5
𝜆 6 , 𝜇 6
𝜆 3 , 𝜇 3
FDEP G1 G2 G3 G4
RFT (λ, μ)
FT(λ, μ=0)
Event probability at G1:
𝐴 𝐺1 = 𝐴 𝐺2 ∗ 𝐴 𝐺3
𝑣 𝐺1 = 1 1/𝑢 𝐺2 + 1/𝑢 𝐺3
𝑢 𝐺1 = 𝑢 𝐺1 ∗(1− 𝐴 𝐺1 ) 𝐴 𝐺1
𝑃 𝐺1 =1− e −( 1 u G1 )t
𝑅 𝐺1 = 𝑅 𝐺2 ∗ 𝑅 𝐺3
𝑃 𝐺1 =1−𝑅 𝐺1
Top event probability
(PG1)
0.06
0.11
G1 & G3 : Series gates
G2 & G3: Parallel gates
λ: failure rate
μ: recovery/repair rate
« Recovery of failed items can reduce up to 54% the chance of top event failure »
Probabilistic Safety Assessment of ASTRID DHR Systems

29. Dynamic Safety Assessment of ASTRID DHR system
PyCATSHOO (Pythonic Object Oriented Hybrid Stochastic Automata) method models a system with interacting probabilistic and deterministic variables explicitly.
The modeling and analysis using PyCATSHOO is currently undergoing on ASTRID DHR system.
PyCATSHOO model of ASTRID DHR systems in which the hybrid automata D1, D2, D3, and POOL communicate real-timely among themselves through message box (coloured) about their state such as random failure (probabilistic) of DHR systems and sodiulm temperature rise (deterministic) in pool
Probabilistic Safety Assessment of ASTRID DHR Systems

30. WP4: Analysis of severe accident scenarios
Postdoctoral researcher:
Diana Caraghiaur
Supervisors:
Pavel Kudinov (KTH)
Nathalie Marie and Frédéric Bertrand (CEA)
WP aimed at assessment of expansion phase of a SFR FCI with simplified models

31. Motivation understanding the phenomena limitation of vessel loadings
Core melting
fuel-coolant interaction
=> coolant vaporisation
power excursion
=> fuel vaporisation
mechanistic modelling
simplified modelling
mechanical energy release due to vapour expansion
understanding the phenomena
limitation of vessel loadings

32. Phenomena of fuel-coolant interaction
Molten fuel-sodium interaction creates favourable conditions for fine fuel fragmentation
This leads to drastic increase of heat transfer surface area, and thus, the amount of heat rapidly transferred from the fuel to the (more volatile) sodium
In consequence, a large amount of sodium vapour is produced in a short time
The specific volume of sodium gas is about 3000 times larger than the specific volume of liquid sodium
The increase in volume produces pressure in the enclosed volume of fuel assembly, reactor core or primary vessel
The mechanical energy is released, which can endanger the surrounding structures
Δ𝑄=ℎ𝐴Δ𝑇
𝐸 𝑀 = 𝑃d𝑉

33. Fine fragmentation – a prerequisite for energetic FCI!
A large subcooling of sodium
The interfacial temperature between molten fuel and sodium can be below the melting temperature of fuel
Fragmentation can be due to formation of a solid crust on the surface of fuel droplet. The crust is ruptured due to an internal pressure build up
Tapez une équation ici.
Illustration of fragments obtained from interaction of a single molten metal droplet penetrating a sodium pool, Zhang et al, 2009
…the largest fragment shows that the inside of the lower hemisphere is empty. The diameter of the hemisphere is almost equal to the initial droplet. The appearance of lower part clearly shows to be the solid crust produced upon contact with sodium..
𝑇 𝑖 =1266 ℃
Illustration of parameters which influence the molten fuel droplet fragmentation due to solidification
𝑇 𝑖 =1004 ℃

34. Example of schematic representation of the system at t=0
Simplified model
Complement to mechanictic tools. The mechanistic tool is not yet available for SFR FCI, thus the simplified model can be the only tool for the design of ASTRID
Fast calculation of parameters of interest – mechanical energy release and pressure evolution
Possibility of conducting large parametric studies
Easy adaptation to various FCI configuration (various scales, design evolution, etc.)
Better treatment of epistemic uncertainties
Consists of:
Heat transfer from fuel to sodium (associated with fragmented fuel)
Energy conservation in the sodium (one- or two-phase)
Equation of state (one- or two-phase)
Constraints (acoustic and inertial)
Example of schematic representation of the system at t=0

35. Model results for different sizes of fuel droplets
𝐸 𝑎𝑣𝑎𝑖𝑙𝑎𝑏𝑙𝑒 = 𝑚 𝑁𝑎 𝐶 𝑝,𝑁𝑎 𝑇 𝑓 − 𝑇 𝑁𝑎
𝑅 𝑓 =143 μ𝑚→ 𝐸 𝑀 =125 kJ→𝜂=3.5%
Geometry used for calculations, applicable to simulate
CORECT 2 experiments. Z(t) represents the interface between the interaction zone and the cold liquid sodium column
𝑅 𝑓 =300 μ𝑚→ 𝐸 𝑀 =74 kJ→𝜂=2.1%
η= 𝐸 𝑀 𝐸 𝑎𝑣𝑎𝑖𝑙𝑎𝑏𝑙𝑒 ∙100%
𝑅 𝑓 =500 μ𝑚→ 𝐸 𝑀 =13 kJ→𝜂=0.4%

36. Future plans Sensitivity studies on fragmentation using CORECT 2 tests
Reactor case application for the ASTRID project at different scales for various typical bounding configurations
Implementation of the tool in the PROCOR CEA severe accident platform

37. Concluding remarks
The ASTRID – safety projects are in progress but it is already visible that this collaboration is quite successful It is not only example of international collaboration - between France and Sweden, but also interdisciplinary one - between safety of LWRs and SFRs