LEADER Project: Task 5.4

Analysis of Representative DBC Events of the ETDR with CATHARE

G. Geffraye, D. Kadri – CEA/GrenobleG. Bandini - ENEA/Bologna

LEADER 5th WP5 MeetingJRC-IET, Petten, 26 February 2013

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Outline

CATHARE code for LFRs CATHARE modelling Steady-state at EOC Analysed DBC transients Transient results Conclusions

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CATHARE code for LFRs

The CATHARE code is the result of a joint effort of CEA, EDF, Framatome ANP and IRSN The CATHARE system code is commonly used for thermal-hydraulic transient analysis and

best-estimate safety analysis of light water reactors. The CATHARE V2.5_2/mod5.1 can be also applied for safety analysis of gas-cooled and sodium-cooled fast reactors

The lead-bismuth eutectic (LBE) and lead thermo-physical properties have been recently implemented in the CATHARE code, in the frame of a bilateral collaboration between ENEA and CEA, for the safety analysis of lead-cooled fast reactors

Some validation works have been performed or are in progress at ENEA on the basis of: The experiments conducted in the Korean LBE-cooled HELIOS loop in the frame of the

OECD LACANES benchmark, including a comparison with RELAP5 code; The experiments carried out in the LBE-cooled NACIE facility of ENEA/Brasimone; The experiments carried out on the LBE-cooled TALL loop of KTH/Stockholm within the

on-going activities of the THINS European project. The whole core, primary system and secondary system discretization of the ALFRED reactor

used in the CATHARE simulation has been harmonized with the one used in the RELAP5 simulation performed by ENEA, in order to better compare the code results

The systematic comparison of CATHARE results with the results of the more validated RELAP5 code, regarding both DBC and DEC transient analysis, has confirmed the good applicability of CATHARE for ALFRED transient simulations

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ALFRED modelling

ALFRED Nodalization scheme with CATHARE

Primary circuit

2 Secondary loops (weight 4)

2 IC loops (weight 4)

Cold pool modelled by two 1D axial elements

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Steady-state at EOC

Parameter Unit CATHAREReactor thermal power MW 300Total primary flowrate kg/s 25740Active core flowrate kg/s 25460Average FA flowrate kg/s 148.7Hottest FA flowrate kg/s 176.8Pressure loss through the primary circuit bar 1.5Pressure loss through the core bar 1.0Core inlet lead temperature °C 400Average FA outlet lead temperature °C 480Hottest FA outlet lead temperature °C 483Upper plenum lead temperature °C 480Average pin max clad temperature °C 508Hottest pin max clad temperature °C 518Average pin max fuel temperature °C 1635Hottest pin max fuel temperature °C 1985SG inlet lead temperature °C 480SG outlet lead temperature °C 400Total SG feedwater flowrate (8 SGs) kg/s 196.6SG feedwater temperature °C 335SG steam outlet temperature °C 451SG inlet pressure bar 188SG outlet pressure bar 182Steam line outlet pressure bar 180

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TRANSIENT Initiating Event (t = 0 s) Reactor scram and threshold

Primary pump trip

MHX FW trip

MSIV closure DHR startup

TD-1: Spurious reactor trip Reactor scram 0 s, Spurious trip No No No No

TD-3: Loss of AC power Station blackout 0 s, CR magnet

de-energization 0 s 0 s 0 s DHR-1 at 1 s (4 IC loops)

TD-7: Loss of all primary pumps All primary pump coastdown 3 s, ΔT hot FA =

120% nominal 0 s 3 s 3 s DHR-1 at 4 s (3 IC loops)

TO-1: Reduction of FW temperature

FW temperature from 335 °C down to 300 °C in 1 s

2 s, Low FW temperature No 2 s 2 s DHR-1 at 3 s

(4 IC loops)

TO-4: Increase of FW flowrate

20% increase in FW flowrate in 25 s

No, No scram threshold reached No No No No

Main events and reactor scram threshold

Analysed DBC transients

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TD-1: Spurious reactor trip (1/2)

Total reactivity and feedbacks

ASSUMPTIONS: Reactor scram at t = 0 s Reactivity insertion of at least 8000 pcm in 1 s Secondary circuits are available constant feedwater flowrate

Core power reduced down to decay level at t = 0 s Power removal by secondary circuits reduces with decreasing primary temperatures

Core and MHX powers

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Core temperatures

Initial temperature gradient on the fuel rod clad is about -8 °C/s No risk for lead freezing since the feedwater temperature (335 °C) remains above

the solidification point of lead (327 °C)

TD-1: Spurious reactor trip (2/2)

Primary lead temperatures

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TD-3: Loss of AC power (1/2)

Active core flowrate

ASSUMPTIONS: At t = 0 s Reactor scram, primary pump coastdown, feedwater and turbine trip At t = 1 s DHR-1 system activation (4 IC loops risk of lead freezing)

Core temperatures

No initial core flowrate undershoot (lead free levels equalization) No significant clad temperature peak in the initial phase of the transient

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TD-3: Loss of AC power (2/2)

Core decay, MHX and IC powers

Primary lead temperatures After the initial transient the natural circulation in the primary circuit stabilizes around 2% of nominal value

DHR power (7 MW) exceeds the decay power after about 15 minutes

After 3 hours the minimum lead temperature at MHX outlet is still far enough from the lead solidification point (mixing in the cold pool around MHXs is effective)

Active core flowrate

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TD-7: Loss of primary pumps (1/2)

ASSUMPTIONS: At t = 0 s All primary pumps coastdown Reactor scram at t = 3 s on second

scram threshold (Hot FA ΔT > 1.2 nominal value) At t = 4 s DHR-1 system activation (3 IC loops maximum temperatures)

Active core flowrate Core temperatures

No initial core flowrate undershoot (lead free levels stabilization) More significant clad temperature peak than in case of LOOP transient due to delayed reactor scram

TD-7: Loss of primary pumps (2/2)

Active core flowrate Core decay, MHX and IC powers

Primary lead temperatures After the initial transient the natural

circulation in the primary circuit stabilizes around 1.5% of nominal value

DHR power (5 MW) exceeds the decay power after about 45 minutes

No risk of lead freezing at MHX outlet in the short and medium term (mixing in the cold pool around MHXs is effective)

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TO-1: FW temper. reduction (1/2)

ASSUMPTIONS: Loss of one preheater (FW temperature from 335 °C down to 300 °C in 1 s)

reactor scram at t = 2 s on low FW temperature At t = 3 s DHR-1 system activation (4 IC loops)

Primary lead temperatures DT through the core and the

MHX reduces quickly down to few degrees

After some fluctuations the primary lead temperatures stabilizes around 425 °C

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Core decay, MHX and IC powers Primary lead temperatures

TO-1: FW temper. reduction (2/2)

No risk of lead freezing in the initial phase of the transient due to prompt reactor scram

After about 15 minutes the DHR power (7 MW) exceeds the decay power No risk of lead freezing in the sort and medium term (mixing in the cold pool

around MHXs is effective)

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TO-1: FW flowrate +20%

Core and MHX powers Primary lead temperatures

ASSUMPTIONS: Feedwater flowrate +20% in 25 s

No significant perturbations on both primary and secondary sides The system reaches a new steady-state condition in about 10 minutes without

exceeding reactor scram set-points Slight increase in core power (+4%) leads to max fuel temperature increase of 40 °C

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Maximum core temperatures

Transient Description Code System Max Temperatures [°C] Fuel Cladding Coolant

Nominal Steady state, peak pin - ENEA CATHARE 1985 518 483

TD-1 Spurious reactor trip CATHARE 1985 518 483

TD-3 Loss of AC power CATHARE 1985 564 539

TD-7 Loss of all primary pumps (PLOF) CATHARE 1985 612 579

TO-1 Reduction of FW temperature CATHARE 1985 518 483

TO-4 Increase of FW flowrate by 20 % CATHARE 2022 518 483

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Conclusions

In all analysed DBC accidental transients the protection system by reactor scram and prompt start-up of the DHR-1 system for core decay heat removal is able to bring the plant in safe conditions in the short and long term. The core temperatures (clad and fuel) always remain well below the safety limits and no significant vessel wall temperature increase is predicted.

The time to reach lead freezing at the MHX outlet after start-up of DHR-1 system strongly depends on the assumptions taken on the lead mixing in the cold pool surrounding the MHX that involves the largest part of the primary lead mass inventory. In the CATHARE calculations the cold lead flowing out of the MHX mixes with hotter lead of the cold pool surrounding the MHX, before to move downward into the lower plenum towards the core inlet. Therefore, in the calculations of TD-1, TD-7 and TO-1 transients, the decrease of lead temperature in the primary system is significantly delayed by the coolant mixing in the cold pool, that increases noticeably the effective thermal inertia of the primary system. This cold pool mixing effect (not observed in the analysis with the RELAP5 code by ENEA) mainly explains the large difference between CATHARE and RELAP5 results, regarding the time needed to approach the risk of lead freezing following DHR-1 start-up.