## Abstract

A large 3600 MW-thermal European sodium fast reactor (ESFR) concept has been studied in a European Horizon-2020 project since September 2017, following an earlier European project. In the paper, we describe new ESFR core safety measures focused on prevention and mitigation of severe accidents. In particular, we propose a new core configuration for reducing the sodium void effect, introduce passive shutdown systems, and implement special paths in the core for facilitation of molten fuel discharge in order to avoid recriticalities after a hypothetical severe accident. We describe and assess the control and shutdown system, and consider options for burning minor actinides.

## 1 Introduction

This paper describes new core safety measures for a 3600 MW-thermal European sodium fast reactor (ESFR) concept, studied since late 2000s in European projects. The first 4-year project was a collaborative project for a European sodium fast reactor (CP-ESFR) [1] on studies of ESFR with both mixed oxide and carbide fuels. The next project since 2017 is a Horizon 2020 project called European sodium fast reactor safety measures assessment and research tools (ESFR-SMART) [2], which continues to investigate the oxide fuel option. We refer to the initial ESFR concept with oxide fuel of CP-ESFR as ESFR working horse (ESFR-WH) in the following.

European sodium fast reactor-WH differs from earlier large European fast core designs appreciably. Due to relatively thick pins, the fuel volume fraction is higher, and the sodium and steel volume fractions are smaller. The fertile blankets are not present. Figure 1 shows the radial core layout, including diverse shutdown devices (DSDs), and control and shutdown devices (CSDs), and a central steel subassembly. A higher fuel volume fraction leads to relatively lower enrichments of 14.6 wt % and 17 wt % in the inner and outer cores, respectively, see pin dimensions and other data for ESFR-WH in Table 1. For the considered fuel cycle with multiple recycling of plutonium, initially from spent light water reactor fuel, the breeding gain (BG) is near zero, marginally positive, and the reactivity loss under irradiation is below 1.5 $per cycle of 410 days, 1$being the effective delayed neutron fraction, that is about 400 per cent mille (pcm) in ESFR. The sodium void reactivity effect (SVRE) is smaller in ESFR-WH than in earlier large European fast reactor designs, in particular due to a lower sodium volume fraction, but definitely positive: about 3 $at the beginning of life (BOL), and 5$ at the end of equilibrium cycle (EOEC). As sodium boiling may occur after a hypothetical severe accident in ESFR, a smaller SVRE is favorable for reactor safety.

Fig. 1
Fig. 1
Close modal
Table 1

Main ESFR-WH parameters

 Number/enrichment of inner core fuel SAs 225/14.6 wt % Number/enrichment of outer core fuel SAs 228/17.0 wt % Number of CSDs 24 Number of DSDs 9 Target fuel residence time (effective power days) 2050 Target burn-up (GWd/t) 100 Fissile core height (cm) 100 HEX SA pitch (mm) 210.8 Fuel pellet diameter (mm) 9.43 Fuel pin outer clad diameter (mm) 10.73 Number of pins per fuel SA 271
 Number/enrichment of inner core fuel SAs 225/14.6 wt % Number/enrichment of outer core fuel SAs 228/17.0 wt % Number of CSDs 24 Number of DSDs 9 Target fuel residence time (effective power days) 2050 Target burn-up (GWd/t) 100 Fissile core height (cm) 100 HEX SA pitch (mm) 210.8 Fuel pellet diameter (mm) 9.43 Fuel pin outer clad diameter (mm) 10.73 Number of pins per fuel SA 271

In Sec. 2, we briefly review core optimization activities previously performed in CP-ESFR. In Sec. 3, we report in more detail—as compared to Ref. [3]—on new ESFR-SMART core safety measures focused on prevention and mitigation of severe accidents. In particular, we describe a new core configuration with a reduced SVRE, implementation of passive shutdown systems and special paths in the core for facilitation of molten fuel discharge. The introduction of these paths may help to avoid recriticalities and therefore to reduce the energy release after a hypothetical accident. We also assess the control and shutdown system in Sec. 4, and briefly consider minor actinide burning options in Sec. 5.

## 2 Collaborative Project-European Sodium Fast Reactor Core Optimization Studies

Preliminary safety assessments did show that relatively large power excursions would be possible in ESFR-WH due to a positive reactivity variation induced by sodium boiling after a hypothetical accident. These excursions could lead to core melting, then to separation of molten steel and fuel, which are materials of different densities. This separation may lead to recriticalities caused by fuel movement [4]. The induced by recriticalities power excursions are challenging events with respect to reactor vessel integrity because of potentially strong reactivity variations due to fuel movement. A final large excursion may happen at some time and force a massive fuel discharge from the core, making then the reactor deeply subcritical, the mechanical energy release caused by this excursion being in the general case higher if more energy is accumulated in the core, mainly due to excursions happened before. In the literature, the transient progression before and shortly after the first excursion—when the intact subassembly (SA) can walls restrict the radial fuel movement—is often referred as the initiation or primary phase. The next transient phase is the transition (to full core melting) or secondary phase.

The core optimization studies in CP-ESFR addressed two issues: (1) reduction of SVRE in order to prevent or limit core damage during the initiation phase and (2) introduction of special paths for facilitation of early molten fuel discharge from the core in order to prevent multiple recriticalities during the transition phase.

### 2.1 Sodium Void Effect Reduction Options of Collaborative Project-European Sodium Fast Reactor.

Several SVRE reduction options were investigated in CP-ESFR, including introduction of moderator materials and ways for making the core flatter, i.e., shortening the core fissile height while increasing the core diameter. A core flattening increases the neutron leakage that is favorable for SVRE reduction, but increases the core radius and the number of SAs that may increase construction and operation costs.

The most effective SVRE reductions proposed in CP-ESFR—that do not change the fissile region geometry—are based on modifications of the axial material arrangements above and below the core. A so-called sodium plenum, i.e., can walls with sodium inside, topped by an absorber layer above, replaces the steel reflector above the core. A similar approach for SVRE reduction is used in the BN-800 reactor as proposed in Ref. [5]. The idea is that sodium boiling in the core would spread to the plenum above; then the neutron leakage through the plenum would increase after its voiding, thus making SVRE less positive. An additional design modification for SVRE reduction is the introduction of a short fertile blanket instead of the steel reflector below the core; this measure facilitates neutron leakage down from the core and further reduces the void effect. The introductions of the plenum with the absorber layer above and of the fertile blanket below the core resulted in a so-called ESFR-CONF2 design, with the same radial arrangement as ESFR-WH, but with a new axial one, see Table 2. The fissile and absorber pins are separated from the sodium plenum by steel plugs, the above-fuel-plugs being shorter than in ESFR-WH in order to make the plenum more effective. SVRE, after voiding the core and plenum in ESFR-CONF2, is about 1 $at BOL, and about 3$ at EOEC, which is appreciably lower than SVRE in ESFR-WH.

Table 2

Axial structure ESFR-CONF2 fuel SA: heights in cm

 Head 23 Reflector 27.6 Absorber 28.2 Plug 1.8 Na plenum 60 Plug 1.8 Upper gas plenum 5 Fissile core 100 Fertile blanket 30 Lower gas plenum 91.3 Plug 8.2 Foot 37
 Head 23 Reflector 27.6 Absorber 28.2 Plug 1.8 Na plenum 60 Plug 1.8 Upper gas plenum 5 Fissile core 100 Fertile blanket 30 Lower gas plenum 91.3 Plug 8.2 Foot 37

### 2.2 Paths for Molten Fuel Discharge in Collaborative Project-European Sodium Fast Reactor.

Also introduction of special SAs for facilitation of molten fuel discharge from the core was considered in CP-ESFR. This approach resembles a fuel SA design option called fuel assembly with inner duct structure (FAIDUS) [6].

In a FAIDUS SA, a duct with sodium inside replaces a group of fuel pins. After a hypothetical accident, molten fuel penetrates into the duct and then discharges through the duct from the core. This approach facilitates early molten fuel discharge within the SA, but makes the SA design more complex.

A modified CP-ESFR design includes, instead of many FAIDUS SAs, few molten fuel discharged tubes, i.e., can walls with sodium inside. These discharge tubes are similar to control rods without absorber, also connected to the inlet sodium plenum. This design is less complex than FAIDUS, but includes stronger obstacles, in particular the SA can walls, for molten fuel relocation. After replacement of 18 fuel SAs by the discharge tubes in the inner core, the outer core contains extra 18 fuel SAs added to the core periphery.

All discharge tubes considered in CP-ESFR are in a hexagonal ring between DSDs and external CSDs. This arrangement aims to prevent or reduce radial inward movement of the outer higher enriched molten fuel, thus decreasing the recriticality potential. The introduction of the tubes is slightly favorable for the sodium void effect: in case of sodium boiling in adjacent fuel SAs, the introduced reactivity effect is slightly smaller than the effect for the configuration with no tubes. This is because sodium remains inside the tubes and moderates neutrons around. A full tube voiding—after a hypothetical accident—results in a negative reactivity effect because of a strong neutron leakage through void tubes.

A limited number of transient simulations for modified designs in CP-ESFR preliminary confirmed [7] a better transient behavior of ESFR-CONF2 compared to ESFR-WH during the initiation phase, but did show that a further SVRE reduction would be of interest. The simulations also confirmed the possibility of molten fuel penetration into the discharge tubes, but simulation results were sensitive to sodium flow rates and temperatures inside the tubes and in the gaps between the tubes and fuel SAs.

Note that transient simulations performed by now relied on rather simplified models for accounting core geometry variations under transient conditions, such as radial core expansion models based on diagrid thermal expansion, axial core expansion models based on fuel or clad expansion. The fuel-driven option is more accurate for nonirradiated fuel. Under irradiation, the gap between fuel and clad may disappear; then the clad-driven model is more accurate. Results that take into account more complex core deformations, such as SA bowing, are not yet available, but the related uncertainties in the total reactivity feedback in the considered large system are less important than in a small reactor.

## 3 European Sodium Fast Reactor-Safety Measures Assessment and Research Tools Core Safety Measures

In ESFR-SMART, we continue studies performed in CP-ESFR, while taking into account core safety measures proposed for a recent design called advanced sodium technological reactor for industrial demonstration (ASTRID) [8].

A 1500 MW-thermal ASTRID sodium-cooled reactor design includes, among others, two particular features. First, the inner fissile region is shorter than the outer one. Shortening of the inner core is more efficient for void effect reduction—per eliminated fissile volume—than shortening of the full core because of a higher neutron leakage from the core periphery. Second, the inner core incorporates an axial fertile blanket at an intermediate axial position. The introduction of this blanket reduces the void effect by increasing the axial leakage of neutrons; it also improves the Pu balance and lessens the reactivity loss under irradiation.

### 4.2 Shutdown Margins.

The reactivity reserve available following a reactor scram from any operational state and considering the stuck rod condition is referred to as the shutdown margin. It constitutes the negative reactivity required to shut down the reactor and to provide reactivity hold-down to maintain subcriticality over a prolonged period with an adequate reactivity margin.

Two sets of calculations were performed at the hot standby temperature (450 °C) in order to estimate the shutdown margins for the cases of reactor scram using CSD/DSD rods and ARG1/ARG2 rods.

The results of the calculations are listed in Table 7. The shutdown margins shown for both shutdown systems appear adequate. The higher reactivity worth of the CSD rods is due to the design requirements, which include a capability to compensate for the excess reactivity and for other reactivity feedbacks including refueling worth uncertainties.

Table 7

Shutdown reactivity margins

Reactor core—safe shutdownShutdown systemReactivity margin (pcm)
ScramCSD−3956
DSD−1359
Control rod failureCSD—1 stuck rod−3567
DSD—1 stuck rod−1249
ScramARG1−2701
ARG2−2712
Control rod failureARG1—1 stuck rod−2570
ARG2—1 stuck rod−2566
Reactor core—safe shutdownShutdown systemReactivity margin (pcm)
ScramCSD−3956
DSD−1359
Control rod failureCSD—1 stuck rod−3567
DSD—1 stuck rod−1249
ScramARG1−2701
ARG2−2712
Control rod failureARG1—1 stuck rod−2570
ARG2—1 stuck rod−2566

The DSD rods have to provide only redundant safety shutdown capability to bring the reactor to zero power at the hot standby temperature from any operation condition considering a stuck rod fault. The reactivity reduction due to the stuck rod fault is for CSD systems in the range of 1 $while for the DSD system it is only 0.35$. For the outer CSD rods, the value is of the order of 1 \$, thus requiring a modification of the absorber material enrichment to adjust the worth distribution between the inner and outer rods of the CSD system.

Results for the ARG1 and ARG2 system show the same level of shutdown margin. The reactivity reduction due to the stuck rod fault in this case is in the range of 130–150 pcm.

### 4.3 Accident Situations.

Considerations related to inadvertent reactivity insertion play a key role in connection with the shutdown systems design requirements. There are various potential reactivity insertion mechanisms, which need to be considered to provide provisions to prevent reactivity fault accidents. For preliminary assessment performed within the present analysis, however, it is sufficient to consider an enveloping insertion mechanism such as the inadvertent withdrawal of the control rod with the highest worth. However, it has to be noted that the control rod drive mechanism has provisions to avoid any inadvertent control rod withdrawal.

The reactivity insertion due to inadvertent withdrawal of the highest worth control rod was estimated by the direct eigenvalue method. The results are shown in Table 8 for both CSD and DSD control rod withdrawal cases.

Table 8

Reactivity insertion due to inadvertent control rod withdrawal

Control rod systemReactivity (pcm)
CSDInner147
Outer362
DSD141
Control rod systemReactivity (pcm)
CSDInner147
Outer362
DSD141

The reactivity insertion due to withdrawal of one inner CSD control rod is in the order of 140 pcm. The DSD control rod withdrawal reactivity insertion is in the same range. What is of concern is the large reactivity insertion due to the withdrawal of outer CSD control rod.

It needs to be further analyzed if this high withdrawal reactivity insertion could be a safety issue. Furthermore, additional calculations are needed to confirm that the calculated reactivity insertions could cause fuel melting and the consequences thereof.

## 5 Options for Minor Actinide Incineration

In the optimized ESFR-SMART design, the Pu balance is near-zero, but the Minor Actinide (MA) one is positive. On the other hand, this design includes a lower blanket that is 25 cm thick below the inner fissile region of 75 cm and is 5 cm thick below the outer one of 95 cm. This blanket can be used for MA incineration without strong influence on core characteristics. One may put in this blanket a mixture of uranium and MA oxides, e.g., with the 80-to-20 ratio, instead of uranium oxide only. Preliminary analyses [19] show that such design modifications may lead to a negative MA balance, while SVRE is slightly reduced because of a higher leakage from the core to the blanket that is a stronger absorber after MA introduction. After this modification, however, the Pu balance becomes more positive. Additional studies would be needed in order to assess in more detail options for MA incineration in the ESFR-SMART and associated fuel cycle.

## 6 Concluding Remarks

A large European sodium fast reactor concept has been studied in European projects since late 2000s. In the initial ESFR design, the fuel pins are relatively thick, the fuel volume fraction is relatively high, the steel and sodium ones are relatively low, that leads to lower enrichments, lower reactivity losses per cycle, and lower sodium void effect reactivity values compared to earlier large European designs. This new design approach supports an optimal utilization of plutonium from spent nuclear fuel. It is also beneficial for reactor safety.

New core safety measures were proposed in CP-ESFR and ESFR-SMART. In particular, a sodium plenum above the core and a fertile blanket below the core were introduced in CP-ESFR for sodium void effect reduction. In addition, the introduction of special tubes for molten fuel discharge in order to prevent multiple recriticality events after a hypothetical severe accident was studied in CP-ESFR.

In ESFR-SMART, we took into account earlier studies and establish a new design. A core with different fissile heights in the inner and outer regions, but with the same upper fissile axial boundary is proposed. The core is optimized with respect to safety and Pu balance by employing a new automated procedure that allows considering a very large number of design options. The fuel enrichment has been fixed to the same value in the inner and outer core regions, offering advantages for fuel fabrication and safety. The void effect was reduced to a near-zero value. On the other hand, the enrichment is higher, and the reactivity loss per cycle is stronger than in the initial ESFR design.

Finally, passive shutdown systems were introduced, and a new arrangement for discharge tubes is employed. The control and shutdown systems assessment was performed, as well as the preliminary assessments on minor actinide incineration.

The new ESFR-SMART design is expected to demonstrate better safety performances. It offers a good basis for ESFR-SMART studies and later projects.

## Funding Data

• Euratom Research and Training Program 2014–2018 (Grant Agreement No. 754501; Funder ID: 10.13039/100010687).

## Nomenclature

• ARG =

absorber rod group

•
• ASTRID =

advanced sodium technological reactor for industrial demonstration

•
• BG =

breeding gain

•
• BOL =

beginning of life

•
• CP-ESFR =

collaborative project for a European sodium fast reactor

•
• CSD =

control and shutdown device

•
• DSD =

diverse shutdown device

•
• EOEC =

end of equilibrium cycle

•
• ESFR =

European Sodium Fast Reactor

•
• ESFR-SMART =

European Sodium fast reactor safety measures assessment and research tools

•
• ESFR-WH =

ESFR working horse

•
• FAIDUS =

fuel assembly with inner duct structure

•
• jeff 3.1 =

joint evaluated fission and fusion nuclear data library version 3.1

•
• MA =

minor actinide

•
• pcm =

per cent mille

•
• SDDS =

•
• SVRE =

sodium void reactivity effect

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