Analysis of Unprotected Transients in the Lead-Cooled ALFRED Reactor

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1 Analysis of Unprotected Transients in the Lead-Cooled ALFRED Reactor G. Bandini (ENEA/Bologna) E. Bubelis, M. Schikorr (KIT/Karlsruhe) A. Alemberti, L. Mansani (Ansaldo Nucleare/Genova) Consultants Meeting: Education Training Seminar on Fast Reactors Science and Technology ITESM Campus Santa Fe, Mexico City - 29 June -03 July 2015

2 Outline Introduction The Lead-Cooled ALFRED Reactor ALFRED Modeling with SIM-LFR and RELAP5 Nominal Steady-State Reactivity Feedbacks Unprotected Transient Results Conclusions 2

3 Introduction The conceptual design of the ALFRED reactor was developed within the LEADER EU FP7 project to meet the safety objectives of GEN-IV nuclear energy systems. One of the main objectives of the project was to perform a preliminary safety analysis of ALFRED. Design Basis Conditions (DBC) and Design Extension Conditions (DEC) have been considered in the safety analysis of ALFRED. The DEC scenarios include the so-called Unprotected Transients characterized by the failure of reactor scram. ALFRED The main objective of Unprotected Transient analysis is to evaluate the impact of the core and plant design features on the intrinsic safety behaviour of the plant. More representative Unprotected Transients for ALFRED have been analysed by various research organizations using different system codes. 3

4 ALFRED primary system Section Reactor block Lead-cooled pool-type reactor of 300 MWth (~ 130 MWe power). Active core fuelled with 171 FAs. Eight Pump-SG groups connected to eight independent secondary circuits. 4

5 ALFRED secondary circuits To DHR system Water In-water pool isolation condenser (IC) Valve Steam Feedwater Steam SG (MHX) From DHR system DHR-1 system (4 IC loops) DHR-2 system (other 4 IC loops) Hot Lead 5

6 SIM-LFR modelling Schematic representation of the primary and part of secondary side of the Pb-cooled ALFRED reactor for the nominal operation power at EOC conditions in SIM-LFR code SIM-LFR code PC-based, multi node point kinetic model: describes both nuclear and thermal-hydraulic transient characteristics of both critical and sub-critical reactor cores. SIM-LFR was validated and used extensively for the transient analysis of the LBEcooled PDS-XADS ADS design (PDS-XADS project), the EFIT-Pb and XT-ADS designs (EUROTRANS project), and critical and sub-critical FASTEF reactor designs (CDT project). 6

7 RELAP5 modelling RELAP5 code Modified Mod3.3 code version with lead and LBE properties implemented by ENEA Ansaldo. Validated on experiments conducted in CHEOPE, NACIE and CIRCE facilities at ENEA/Brasimone. Used in EFIT, ELSY and MYRRHA safety analysis. Nodalization scheme of the primary system, secondary circuits and DHR loop in RELAP5 The active core is represented by: 1 average FA representing 170 FAs 1 hot FA including the peak power pin 7

8 Nominal steady-state Parameter Unit ALFRED RELAP5 SIM-LFR Reactor thermal power MW Total primary flow rate kg/s Total ΔP in the primary circuit bar ΔP through the core bar < Core inlet temperature C Upper plenum temperature C Max core outlet temperature (*) C Peak clad temperature C ~ Peak fuel temperature C ~ Feedwater temperature C Feedwater flow rate kg/s Steam temperature C Steam pressure bar (*) Hottest FA mass flow rate is ~120% of average FA mass flow rate. 8

9 Reactivity feedbacks REACTIVITY COEFFICIENT Unit Ref. Temperature Value Control rod differential expansion (*) pcm/k T upper plenum Coolant expansion (**) pcm/k Average T-core Axial clad expansion pcm/k Average T-clad Axial wrapper tube expansion pcm/k Average T-wrapper Radial clad expansion pcm/k Average T-clad Radial wrapper tube expansion pcm/k Average T-wrapper Diagrid radial core expansion pcm/k T-core inlet Pad radial core expansion pcm/k T-core outlet Axial fuel expansion: free pcm/k Average T-fuel Axial fuel expansion: linked pcm/k Average T-clad Doppler constant pcm Average T-fuel -566 (*) Prompt response (the delayed response has been neglected). (**) Calculated on the whole height of the fuel assembly (the other feedbacks are calculated only in the fissile zone). 9

10 Transient Analysis Transient Initiating Event CODE UTOP Reactivity insertion of 250 pcm in 10 s (core compaction or core voiding following SGTR) SIM-LFR ULOF Loss of all primary pumps SIM-LFR ULOHS Loss of feed water to all steam generators RELAP5 ULOHS+ULOF OBJECTIVES: Verify the intrinsic safety behavior of the ALFRED plant and its response to more unlikely accidental events. Verify the safety limits for the fuel rod clad and the vessel wall structure. Verify the grace time available to operator for corrective actions. Loss of feed water to all steam generators + loss of all primary pumps RELAP5 10

11 rel. units [fr] Temperature [ C] UTOP: SIM-LFR Results 250 pcm reactivity insertion within 10 s time interval at HFP conditions. Normal reactor operation at hot full power (HFP) conditions was assumed. Relative core thermal power and flow rate Power_th Flow_Cool Time [sec] Peak fuel-, clad-, coolant- and vessel wall temperatures Fuelc_peak Cool_out T_wall Clad_peak Cool_in Time [sec] Insertion of 250 pcm in 10 s time interval at the end of equilibrium cycle (EOC) conditions leads to a power jump of ~ 2.2 nominal, assuming reactor shut-down system failure. Max. fuel and clad temp. increase: 2064 C 2996 C and 514 C 631 C, respectively, resulting in short-term (~ 100 s) local fuel melting (fuel center) of the peak fuel pins. 11

12 Reactivities [pcm] UTOP: SIM-LFR Results Total reactivity and feedbacks Time [sec] Doppler Diagrid Exp Fuel Expan Rod Drive Exp Coolant Exp TotExp:(rods+grid+cool) Rods+step+rate Clad Burn Total The positive reactivity introduced is counterbalanced by negative reactivity from Doppler, fuel, diagrid and coolant expansion effects. ALFRED reactor peak fuel pin cladding accommodates this transient, however localized fuel melting might be encountered in the center of the peak fuel pins. 12

13 rel. units [fr] Temperature [ C] ULOF: SIM-LFR Results All primary pumps and reactor shut-down systems are assumed to fail. Secondary heat transport system is assumed fully functional Relative core thermal power and flow rate Power_th Flow_Cool ~ 60% nominal ~ 24% nominal Time [sec] Peak fuel-, clad-, coolant- and vessel wall temperatures T-fuel-max Fuelc_peak Cool_out T_wall Clad_peak Cool_in Time [sec] Coolant flow drops down to ~ 19 % nominal flow ~ 6 s after the transient initiation, and then recovers to the natural convection asymptotic flow fraction of ~ 24 % at ~ 60 s. Clad temperature of the peak pin (EOC conditions) spikes from 514 C up to max. 751 C at transient time t = 17.8 s due to the coolant flow undershoot. 13

14 Reactivities [pcm] ULOF: SIM-LFR Results Total reactivity and feedbacks Time [sec] Doppler Diagrid Exp Fuel Expan Rod Drive Exp Coolant Exp TotExp:(rods+grid+cool) Rods+step+rate Clad Burn Total The positive reactivity introduced by Doppler and fuel expansion is counterbalanced by negative reactivity from diagrid, CR drive and coolant expansion effects. ULOF transient in the ALFRED can be accommodated, implying no clad failures are expected under minimum coolant flow (flow undershoot) conditions. ULOF transient is thus not a critical transient for the ALFRED reactor. 14

15 ULOHS: RELAP5 Results Loss of feed water to all SGs (MHX) at t = 0 s. Startup of DHR-1 system (only 3 out of 4 IC loops are supposed to be functional). Core and SG (MHX) powers Core and vessel temperatures T-max clad T-max vessel Core power SG power T-core in-out Core power reduces down near the decay level due to negative feedbacks with primary temperature increase core power is removed by DHR-1 in the medium term. Maximum clad and vessel temperatures rise up around 700 C in about 1 hour. 15

16 ULOHS: RELAP5 Results Core temperatures Total reactivity and feedbacks Doppler T-max fuel Fuel exp. T-core inlet T-max clad C.R. exp. Core rad. exp. Cool. exp. The positive reactivity introduced mainly by Doppler and fuel expansion effects is counterbalanced by negative reactivity from core radial, coolant and CR drive exp. effects. No clad failure is expected in the short and long term. Vessel integrity is guaranteed in the medium term but not in the long term. 16

17 ULOHS+ULOF: RELAP5 Results Loss of feed water to all SGs (MHX) and all primary pump coastdown at t = 0 s. Startup of DHR-1 system (only 3 out of 4 IC loops are supposed to be functional). Active core mass flow rate Core temperatures T-max clad T-core outlet Core flow rate T-core inlet The first part of the transient is characterized by the loss of primary pumps with fast transition from forced to natural circulation in the primary system. Clad temperature of the peak pin (EOC conditions) spikes from 508 C up to max. 772 C at transient time t = 11 s due to the initial coolant flow undershoot. 17

18 ULOHS+ULOF: RELAP5 Results The second part of the transient is characterized by the loss of heat sink with progressive primary system heatup and consequent core power reduction due to negative feedbacks. Active core mass flow rate Core and SG (MHX) powers Core power Core flow rate SG power Natural circulation in the primary circuit reduces down to very low value (around 1%). As under ULOHS, the core power reduces down near the decay level until it is removed by the DHR-1 system in the medium term. 18

19 ULOHS+ULOF: RELAP5 Results Core temperatures Total reactivity and feedbacks T-max clad T-core outlet Doppler Fuel exp. T-core inlet T-max vessel Core rad. exp. Cool. exp. C.R. exp. Maximum clad temperature stabilizes around 820 C after about 20 min, while maximum vessel temperature remains well below the safety limit of 550 C for DBC. Similar reactivity feedback behavior observed under ULOHS transient. Clad integrity is not guaranteed in the long term (min t-fail of 3 h predicted by SIM-LFR). No vessel failure is expected in the medium and long term. 19

20 Conclusions (1/2) UTOP Transient: Peak fuel pin cladding is not expected to fail, since there is a large margin to creep rupture limit. The more critical issue is the localized maximum fuel temperature. Fuel melting is observed only in the center of the peak fuel pin (pellet). Extended fuel rod damage can be excluded, as well as the likelihood for this transient to evolve towards a severe accident. ULOF Transient: No relevant flow undershoot is expected in the initial phase of the transient. Stable natural circulation establishes in the primary circuit. The prevalent negative reactivity feedbacks reduce the core power down to about two-thirds of the nominal value. The ULOF transient can be easily accommodated since the maximum clad temperature remains well below the creep rupture limit in the long term. 20

21 Conclusions (2/2) ULOHS Transient: The maximum clad temperature is not a critical issue. The main concern is on the average vessel temperature the structural integrity of the vessel is questionable in the long term. Sufficient grace time (>> 30 min) is available for corrective operator action. ULOHS+ULOF Transient: Due to very low natural circulation flow rate the maximum clad temperature is about C higher than under ULOHS. There is no safety concern on the vessel wall temperature. Sufficient grace time (>> 30 min) is available for corrective operator action. The analysis of unprotected transients has demonstrated the good inherent safety features of the core and the very robust nature of the ALFRED design. In all investigated cases, the large grace time allows the operator to take opportune corrective action for manually shutting down the reactor and then to cool down the primary circuit towards the cold safe shut-down conditions. 21

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