1. Analyses of representative DEC events of the ETDR
LEADER WP5 MEETING, Petten – 26th of February 2013
Analyses of representative DEC events of the ETDR
Kaspar Kööp, Marti Jeltsov
Division of Nuclear Power Safety
Royal Institute of Technology (KTH) Stockholm, Sweden

2. ETDR – ALFRED description
Pool-type
300 MWth
Core pressure drop 1 bar
Temperature
Core inlet 400 C
Core outlet 480 C
Coolant velocity
Average 2 m/s
Maximum 3 m/s
Lead void effect at EOC (only the fuel zones)
+2 $

3. KTH contribution T-DEC1 – complete loss of forced flow + SCRAM fail
T-DEC4 – complete loss of forced flow, complete loss of SCS, DHR system operating

4. T-DEC1&4 RELAP5 model

5. T-DEC1 – Description
T-DEC1 – complete loss of forced flow + SCRAM fail
Pumps are tripped at 0s
Secondary side is operational, IC valves closed

6. T-DEC1 - loss of 8 pumps

7. T-DEC1 - loss of 8 pumps

8. T-DEC1 - loss of 8 pumps

9. T-DEC1 - loss of 8 pumps

10. T-DEC4 – Description
T-DEC4 – complete loss of forced flow + complete loss of secondary cooling system + SCRAM fail
Pumps and SCS are tripped at 0s, IC valves opened at 1s

11. T-DEC4 – loss of flow + loss of SCS + IC valves open

12. T-DEC4 – loss of flow + loss of SCS + IC valves open

13. T-DEC4 – loss of flow + loss of SCS + IC valves open

14. T-DEC4 – loss of flow + loss of SCS + IC valves open

15. T-DEC1&4 RELAP5 model

16. T-DEC4 with 1 working pump

17. T-DEC4 with 1 working pump

18. T-DEC4 with 1 working pump

19. KTH contribution
TR-4 – a transient event due to reactivity insertion (enveloping SGTR, flow blockage, core compaction)

20. TR-4 – Description
TR-4 – a transient event due to reactivity insertion (enveloping SGTR, flow blockage, core compaction)
Steam Generator Tube Leakage (SGTL) is assumed to be the cause for reactivity insertion (voiding of part of active region
We address the task on the transport of bubbles that have leaked in the SG to the primary coolant flow
Reactor is at hot full power (HFP)
Actions:
First thermal-hydraulic part (CFD analysis of bubble transport)
Neutronic part (SERPENT code to look at the consequences of different local core voiding that are typical for SG leakage)

21. ALFRED design
ALFRED – Advanced Lead Fast Reactor European Demonstrator
Power – 125 MWel (300 MWth, ~41% efficiency)
8 steam generators with 8 axial pumps
2 independent DHRs using SG tubes

22. Motivation Core – submerged in the bottom middle part of the pool
8 SGs – located inside the primary circuit radially around the core
One co-axial pump per SG
High P secondary vs low P primary system
SG tube rupture and leakage events in PWRs suggest that it can be a concern in LFRs
Risk of core voiding in case of SGTL due to:
Proximity of SGs to core
Nature (shape) of the flow path
Leak-Before-Break (LBB) (< liter/day in PWRs)
SGTL can hinder licensing due to:
Potential severe consequences (core damage)
Lack of operational experience (uncertainty in frequency of SGTL)
Secondary circuit removed, that is in place in SFRs.
Secondary side pressure is 180 bar.
For compactness, SGs are put into primary pool.
Consequences, core voiding= reactivity insertion, local dry out (burnout) of a fuel rod, rapid transient void insertion.
LFRs have never been built (except military submarines).
9 ruptures in US between 1975 and 2000 and 40 leakage cases in US between 1990 and 1998.
Lead is MORE corrosive and has higher density.
20 MPa in secondary side, whereas only hydrostatic pressure in the primary pool.

23. Objectives To assess the risk related to SGTL
To identify the scenarios of core voiding
To quantify the likelyhood that a steam bubble is transported from a SG to the core
Identify the uncertainties in SGTL
Quantifiy these uncertainties
To estimate void accumulation rates in the core
To estimate the consequent effect to power (or to neutron flux) with a neutronic code
: 9 SGTR and : 40 SGTL incidents
SG in close proximity to core and directo contact to liquid lead.
20MPa in the secondary side

24. Scenarios SGT leakage is considered during normal operation
If steam bubbles are dragged to the core then there are 3 distinct scenarios of safety concern:
Homogeneous voiding of the coolant (continuous leak, small bubbles)
Bubbles stuck in spacers (mid-size bubbles)
Slugs of void entering the core (formed in stagnation zones)
Depending on the scenario...
RIA
Local damage (burn-out) of the fuel
Overpressurization of the vessel
...can happen.

25. Uncertainties
Aleatory and epistemic uncertainties assessed in the SGTL scenarios and phenomenology:
Crack size and morphology
Bubble size distribution
Leak rate through the crack
Bubble drag correlation
Leak rate depends on crack size, morphology and pressure differences between two sides
”Leak-before-break”
K. Terasaka et al. (2011)
A. V. Beznosov et al. (2005)

26. Drag coefficient
Non-linear drag coefficientm, 𝐶 𝐷 , as a function of bubble diameter
Validation is done by comparing 𝑣 𝑇 with predictions from Stokes and Mendelsons law
Bubbles were modeled:
As Lagrangian particles
Constant density
Drag coefficient
With/without turbulent dispersion
Tomiyama et al. correlation chosen:
C D =max min 16 R e B R e B , 48 R e B , 8 3 Eo Eo+4

27. Approach to estimate core voiding
Hot free surface
Cold free surface 2 1 3
Approach to estimate core voiding
Bubbles are injected at
- 3 different height levels in SG (bottom, middle, top)
- at the exit of the core
- at the pump outlet
to estimate probabilities:
P1 – that bubble enters core
P2 – that bubble proceeds to pump
P3 – that bubble stays in the primary loop
Steam accumulation rate in the primary system:
and in the core (assumed bubbles get stuck there):

28. ALFRED CAD model

29. ALFRED CFD model 45° CFD model is being created
Simplified modeling of complex (SG, lower/upper grids) components (porous media, momentum source etc)
Consists of 7 regions a
Downcomer
Lower grid
Lower inactive region
Active region
Upper inactive region
Pump channel
Steam generator

30. Example of ELSY modeling
Core SG
Down-
comer
Pump

31. Example of ELSY results
dbubble=0.2 mm
dbubble=0.4 mm
dbubble=0.5 mm
dbubble=1.0 mm
Very small bubbles are dragged to core (<0.4 mm), whereas middle size ones are not

32. Steam generator tube leakage accident is addressed
Summary
Steam generator tube leakage accident is addressed
Motivation, scenarios, uncertainties
Nominal operational primary flow conditions will be modeled with a CFD code Star-CCM+
Neutronics part of the analysis will be done using Serpent Monte Carlo code
Input for neutronic calculation
void characteristics:
accumulation rates
voiding scenarios are input for neutronics calculation
geometry
ALFRED model exists in the house

33. Thank you!