# Decay Heat Removal and Transient Analysis in  Accidental Conditions in the EFIT Reactor

Authors: Giacomino Bandini; Paride Meloni; Massimiliano Polidori; Maddalena Casamirra; Francesco Castiglia; Mariarosa Giardina
Journal: Science and Technology of Nuclear Installations
 (2008)

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## Abstract

The development of a conceptual design of an industrial-scale transmutation facility (EFIT) of several 100 MW thermal power based on accelerator-driven system (ADS) is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related decay heat removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which are caused by a loss-of-heat sink (LOHS). In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios.


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## Body

## 1. Introduction

Within the  EURATOM Sixth Framework
Program (FP6), the EUROTRANS integrated project [1] is expected to provide a
significant contribution to the demonstration of the industrial transmutation
through the accelerator-driven system route. The goal will be reached through
two phases: the realization of a detailed design for an experimental facility
of 50 to 100 MWth power which shows the technical feasibility of transmutation
in an ADS (XT-ADS), and at the same time, the development of a conceptual design
of a generic European Transmutation Demonstrator (ETD) of several hundreds of
MWth, to be realized in the long term (EFIT).The EFIT reactor should be able to
produce energy at reasonable costs with enhanced transmutation performances,
maintaining as much as possible the high safety level. Modifications introduced
in the plant layout with respect to smaller facilities contribute for a more
compact primary system and a higher core power density: the elimination of
intermediate loops by installation of steam generators inside the primary
vessel and the implementation of mechanical pumps for forced circulation.In order to assure a high safety level, a DHR
system provided with four independent inherently safe loops is installed in the
primary vessel to remove the decay heat by natural convection circulation under
accidental conditions which are caused by a loss of heat removal by the
secondary side through the steam generators.In the present
study, performed in the frame of a collaboration between ENEA and the University of Palermo, the multi-D SIMMER-III code has
been applied to confirm the adequacy of the DHR system design and for the
calibration of the 1D RELAP5 model to be used for T/H transient analysis of
EFIT.


## 2. The EFIT Reactor

The EFIT reactor
[2] is a pool-type reactor which uses pure melted lead as primary coolant in
forced circulation by means of 4 mechanical pumps placed in the hot collector
zone. The reactor thermal power, which is approximately 400 MW, is removed in
normal operation by 8 helical-coil tube bundle steam generators, which are
located in the upper part of the primary vessel to enhance natural circulation
in case of loss of primary pumps. The scheme of the EFIT reactor block is
depicted in Figure 1.Figure 1
Scheme of the EFIT reactor block.The main
thermal-hydraulic parameters of the EFIT reactor are given in Table1. The core
power is removed by 32.3 tons of lead working between 673 K and 753 K.  The
primary coolant at the core outlet is sucked by pumps straight inside the steam
generators, then comes out and flows towards the vessel bottom for coming back and cooling the core. So the
vessel will be in contact with the coolant at its minimum temperature (673 K),
avoiding mechanical stresses due to high lead temperature at core outlet.Table 1
Main EFIT thermal-hydraulic
parameters.
Reactor power (MW)384Core lead mass flow rate (kg/s)32300Core lead temperature (inlet/outlet, K)673/753Primary circuit total pressure drop (MPa)About 0.08Core pressure drop (MPa)About 0.04Secondary steam pressure (MPa)14Secondary steam temperature (K)723Feedwater temperature (K)608DHR system power (4 units, MW)26.6In the  upper
annular space between the inner cylindrical vessel and the main reactor vessel,
the heat exchangers of the DHR system are placed for core decay heat removal
under accidental conditions.


### 2.1. The Decay Heat
Removal System

The DHR system is
conceived for inherently safe decay heat removal in accidental conditions by
means of natural circulation and with passive mode actuation. The system
consists of 4 independent loops filled with organic diathermic fluid (oil) that
dissipate the decay heat to the atmosphere by natural convection circulation.
The functional scheme of one DHR loop is depicted in Figure2. Each loop
consists of a dip cooler (decay heat eXchanger, (DHX)) immersed in the lead,
where the oil partially vaporizes (oil boiling point determined by superimposed
pressure of inert gas), and an air-vapour condenser with stack chimney and
interconnecting piping. The lead enters radially the DHX through a window in
the cylindrical shell and leaves axially the DHX through the open bottom end of
the shell.Figure 2
Functional scheme of one DHR loop.At normal
operating conditions, the oil is below its boiling point and the DHR system
removes the heat losses from steam generators and inner vessel (a few 100 kW)
to keep cold the upper part of the reactor vessel. In accidental conditions
(e.g., LOHS), when the lead temperature increases in the annular space where
the DHXs are located, the oil starts to boil enhancing heat transfer in the DHR
and thus favouring natural circulation on both primary and secondary sides.
Each DHX is rated at approximately 6.7 MW in decay heat removal conditions.


## 2.1. The Decay Heat
Removal System

The DHR system is
conceived for inherently safe decay heat removal in accidental conditions by
means of natural circulation and with passive mode actuation. The system
consists of 4 independent loops filled with organic diathermic fluid (oil) that
dissipate the decay heat to the atmosphere by natural convection circulation.
The functional scheme of one DHR loop is depicted in Figure2. Each loop
consists of a dip cooler (decay heat eXchanger, (DHX)) immersed in the lead,
where the oil partially vaporizes (oil boiling point determined by superimposed
pressure of inert gas), and an air-vapour condenser with stack chimney and
interconnecting piping. The lead enters radially the DHX through a window in
the cylindrical shell and leaves axially the DHX through the open bottom end of
the shell.Figure 2
Functional scheme of one DHR loop.At normal
operating conditions, the oil is below its boiling point and the DHR system
removes the heat losses from steam generators and inner vessel (a few 100 kW)
to keep cold the upper part of the reactor vessel. In accidental conditions
(e.g., LOHS), when the lead temperature increases in the annular space where
the DHXs are located, the oil starts to boil enhancing heat transfer in the DHR
and thus favouring natural circulation on both primary and secondary sides.
Each DHX is rated at approximately 6.7 MW in decay heat removal conditions.


## 3. Simulation of DHR Operation with the SIMMER-III Code

The SIMMER-III
code [3], jointly developed by JNC (J), FZK (D), and CEA (F), is an advanced
safety analysis computer code originally developed to investigate postulated
core disruptive accidents in liquid-metal fast reactors. SIMMER-III is a
two-dimensional, three-velocity field, multiphase, multicomponent, Eulerian,
fluid-dynamics code coupled with a space-dependent neutron kinetics model. By
integrating all the original physical models, SIMMER-III is now applicable to a
large variety of reactor calculations and other complex multiphase flow problems.
The code is provided with a turbulent diffusion model to evaluate the effects
of recirculating flows. This model, which is simpler than classical turbulent
models of CFD codes, has been applied in the present analysis.The EFIT reactor
has been modelled in 2D cylindrical geometry with SIMMER-III (see Figure3).
The core is represented by 6 radial fuel rod rings plus the reflector and
bypass zones, and discretized in 14 axial nodes. The primary pump section, the
steam generators, and the DHR heat exchangers are represented by annular zones
with equivalent cross flow area. The simulation of DHR loops is limited to the
in-vessel heat exchangers, and the power removal is calculated as a function of
the temperature difference between the primary flowing lead and the boiling oil
in the secondary side. As a conservative assumption in the accident analysis,
the DHR system is considered in degraded conditions with 3 out of 4 loops in
service.Figure 3
SIMMER-III nodalization scheme of
the EFIT reactor.A protected LOHS
scenario has been simulated with the SIMMER-III code in order to evaluate the
capability of the DHR system for decay heat removal in transient conditions by
natural circulation in the primary system. Stagnant lead is assumed inside the
primary vessel at transient initiation with simultaneous loss of primary pumps
and steam generator heat removal function. Initial lead and core temperatures
(clad and fuel) are set according to RELAP5 steady-state results for reactor
operation at 384 MW nominal power.The distribution
of lead temperature in the primary vessel calculated by SIMMER-III is
represented in Figure4 from the beginning up to 1 hour transient. Initially,
the release of heat from hot fuel rods leads to lead heatup and strong
recirculation in the upper plenum. Due to the different density and gravity effect,
hot lead flowing down at the steam generator outlet moves upward in the annular
external region between the inner cylindrical vessel and the reactor vessel
where the DHR heat exchangers are located. A natural circulation then starts
inside the DHX where the lead is cooled by efficient heat transfer to the
boiling oil of the secondary circuit. A mixing region of cold and hot fluids is
evidenced below the DHX and the steam generator resulting in temperature
increase of the cold lead flowing down at the DHX outlet towards the core
inlet. Enhanced temperature stratification is evidenced after 1 hour transient
when quasi steady-state conditions are reached, and the core decay power is
efficiently removed by natural circulation through the core and the DHR system
with limited lead temperature increase.Time evolution of lead temperature
(K) distribution in the primary vessel.
(a)(b)(c)(d)(e)(f)At present, the
SIMMER-III code is not validated for this kind of analysis. The results of
lead-bismuth natural circulation experiments, which are foreseen in the
integral CIRCE facility at ENEA/Brasimone research centre, could be used to
confirm the capabilities of the code in this area and propose possible and
beneficial code model improvements.


## 4. Calibration of the RELAP5 Model on SIMMER-III Results

The RELAP5 code [4],
developed by INEEL for the US-NRC,
is a lumped parameter code used for best-estimate thermal-hydraulic (T/H)
transient analysis in light water reactors. The RELAP5 code has been modified
by ANSALDO (including lead properties) to be used for lead-cooled ADS analysis.
Lead
thermophysical property data and heat transfer correlations for heavy liquid
metal in different geometries have been taken from the literature and available
studies and introduced in the modified code [5]. Conservative assumptions have
been considered for heat transfer in core bundle geometry, and new correlations
have been developed for helical tube geometry of SGs [6]. The so modified
RELAP5 code has been applied in previous ADS plant transient analysis such as
PDS-XADS [7], and the results have been successfully compared with other codes.
Furthermore, the code has been validated against experimental data from tests
conducted on the CHEOPE and CIRCE facilities at ENEA/Brasimone Centre [8, 9]; a
new experimental program to be conducted on the large-scale integral facility
CIRCE will provide further valuable data for code validation and verification. This modified version of RELAP5 code is used in the framework of a
collaboration between ENEA Centre and the Department of Nuclear Engineering
(DIN) of the University of Palermo for T/H
transient analysis of EFIT.The  RELAP5 nodalization
scheme of the EFIT reactor employed for T/H transient analysis is shown in
Figure5. In the model, the lead mass inventory distribution and the major flow paths are
represented. The core is discretized in
three zones, each one represented with average and hot channel with equivalent
flow area coupled with corresponding pin thermal structures. The reflector and
bypass zone is represented by an equivalent hydraulic channel, while the target
unit is not simulated. Primary pumps are modelled according to preliminary EFIT
pump design. Steam generator primary (shell) and secondary side (straight and
helical tubes) are modelled in details, while just boundary conditions are used
for the secondary loop. The DHR system model is limited to the DHX primary side
imposing the exchanged power as a function of the lead temperature at the DHX
inlet.Figure 5
RELAP5 nodalization scheme of EFIT
reactor.In case of loss of primary
pumps and LOHS scenario with operation of the DHR system, the natural
circulation flow paths in the 1D RELAP5 representation of EFIT are arbitrarily
defined in the input nodalization scheme. Therefore, mixing effects at steam
generator and DHX outlets and inside plenum recirculation phenomena, observed
with more detailed SIMMER-III analysis, are not taken into account. These
phenomena may tend to reduce the lead mass flow rate in natural circulation
through the core and the DHR system in accidental conditions, as demonstrated
by comparing SIMMER-III and RELAP5 results under the same transient conditions.
As expected, RELAP5 predicts larger lead mass flow rate than SIMMER-III through
the core and the DHR system.The RELAP5 model has been
calibrated in order to reproduce as close as possible the SIMMER-III results. According
to nomenclature in Figure5 (green characters), fluid mixing in the volume
below steam generator and DHX is defined by the fraction of lead mass flow rate
(x=17%) entering this volume from the DHX outlet. This fraction is deduced
from SIMMER-III results by mass and energy balances for the volume according to(1)y=mC(TCi−TDo)(TDi−TDo),x=y+mD−mC, where  C and D denote,
respectively, core and DHX parameters (inlet/outlet temperatures and mass flow
rate of lead).Additional
pressure drop coefficients have been implemented in the revised RELAP5 model of
the primary system (at pump and DHX outlet locations) to reproduce the natural
circulation mass flow rates calculated by the SIMMER-III code in the medium term.The lead mass flow
rate and decay heat removal in the DHR system calculated by the two codes are
presented in Figure6. Both codes predict efficient removal of decay power
after about 2000 seconds. Mass flow rates through the core and the DHR heat exchangers
are equivalent in the medium term. Code
results differ in the initial transient owing to the different modelling.Lead mass flow rates (core and
DHR) and heat removal in the DHR system (3 units).
(a)(b)


## 5. LOHS Accident Analysis with RELAP5

The LOHS
accidental transient has been analysed with the revised RELAP5 model. The
initiating event is the total loss of feedwater of secondary loops. The
following lead temperature increase in the primary circuit is supposed to lead
to reactor trip on high core outlet temperature signal by the protection
system. Reactor trip, consisting in the proton beam switch-off, is assumed with
1 second delay after the average core outlet temperature exceeds 773 K (20 K
above the nominal outlet temperature). As a conservative assumption, primary
pumps trip is assumed to occur at the same time as the reactor trip and no pump
inertia is considered (pump design is not precisely defined yet), in order to
maximize core peak temperature just after reactor trip.The results of the
analysis are presented in Figures7 and 8. Reactor trip is calculated by RELAP5
at 46 seconds. After some initial oscillations induced by free level movements
(see Figure 7), the lead mass flow rates through the core and the DHR system
become stable and the DHR attains maximum performance (20 MW for the 3 units
supposed to be in operation) after about 700 seconds.Lead mass flow rates (core and
DHR) and heat removal in the DHR system (3 units).
(a)(b)Maximum core (lead, clad, and
fuel) and vessel wall temperatures.
(a)(b)The peak  clad temperature reaches 862 K in the
hottest channel of outer  core zone as shown in Figure8 (the clad safety limit
in normal operation of 823 K is exceeded for few seconds only), then all
temperatures stabilize starting from 5000 seconds at acceptable values. The
reactor vessel wall temperature reaches a maximum value of 724 K around 2200 s
during the transient, then it stabilizes at 713 K in the medium term, below the
maximum acceptable value of 723 K. This vessel temperature limit has been
defined to assure the vessel integrity for the entire life of the plant. The
vessel wall temperature peak of 724 K is not a critical issue, because its
duration is limited in time. Furthermore, an improved DHX solution now
implemented in the EFIT design, with increased length of the external wall and
reduced pressure drops, which facilitate the natural circulation of lead
through the DHX, foresees an increase of about 15% of the lead mass flow rate,
thus reducing significantly the thermal load on the vessel wall.The flow path imposed in the RELAP5 model
accelerates the startup of natural circulation through the DHX with respect to
the SIMMER-III simulation (see comparison of mass flow rates calculated by the
two codes in Figure6); however, the integral power removed by the DHR system
at 2200 seconds calculated by SIMMER-III is consistent with the RELAP5 value,
therefore, no appreciable increase of the peak vessel wall temperature is
exhibited at that time by SIMMER-III with respect to RELAP5 analysis.


## 6. Conclusions

The performances
of the DHR system provided in the EFIT reactor have been confirmed by SIMMER-III
and RELAP5 analyses of accidental conditions with complete loss of heat removal
by the secondary system. In particular, natural circulation in the primary
circuit through the core and the DHR system stabilises in less than one hour,
and three out of four DHR units are sufficient to adequately remove the core
decay power under LOHS transient conditions.The 1D RELAP5
model of EFIT has been successfully calibrated on the basis of the SIMMER-III
results by evaluating the amplitude of hot and cold lead mixing at steam
generator and DHX outlets and the effects of turbulence phenomena and
recirculating flows in the upper and lower plena of the reactor vessel. Except
for the initial transient, the results obtained with the revised RELAP5 model
are close to the SIMMER-III results. However, code validation for this kind of
applications is still in progress, and limited for SIMMER-III code. Therefore,
the results of the DHR system performance analysis will be precised after
further validation of the codes at the CIRCE facility or other appropriate
facilities.Finally, the
application of RELAP5 to the LOHS accident analysis has shown that the DHR system
is able to face up accidental situations with sudden total loss of heat removal
by the secondary side with limited increase in core temperature and brings the
plant in safe conditions in the medium term. Also the vessel wall temperature
increase is limited below acceptable value in the medium term with adequate
margin, taking into account more recent DHX design improvements not addressed
in this study.



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