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

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 $

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

T-DEC1&4 RELAP5 model

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

T-DEC1 - loss of 8 pumps

T-DEC1 - loss of 8 pumps

T-DEC1 - loss of 8 pumps

T-DEC1 - loss of 8 pumps

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

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

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

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

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

T-DEC1&4 RELAP5 model

T-DEC4 with 1 working pump

T-DEC4 with 1 working pump

T-DEC4 with 1 working pump

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

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)

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

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.

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 1975-2000: 9 SGTR and 1990-1998: 40 SGTL incidents SG in close proximity to core and directo contact to liquid lead. 20MPa in the secondary side

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.

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)

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 1+0.15 R e B 0.687 , 48 R e B , 8 3 Eo Eo+4

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):

ALFRED CAD model

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

Example of ELSY modeling Core SG Down- comer Pump

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

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

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