Final Report Summary - CP-ESFR (Collaborative Project on European Sodium Fast Reactor)
Executive Summary:
The key research goals for fourth generation of European sodium cooled fast reactors can be summarized as follows:
• An improved safety.
• A financial risk comparable to that of the other means of energy production.
• A flexible and robust management of the nuclear materials.
All these issues were addressed by the project. In addition, it contributed to rebuilt the European expertise on sodium cooled fast reactors
The project was organized in 6 Sub projects:
- SP0: Management (administrative & technical activities)
- SP1: Consistency and assessment
- SP2: Fuel, fuel element, core & fuel cycle
- SP3: Safety concept options and PR & PP issues
- SP4: Innovative Reactor Architecture, components and BOP
- SP5: Education & training
Main results of the project
The main results since the beginning of the project are:
- Detailed definition of starting points for the studies (working horses) for the core and the reactor.
- The identification of core design options to improve inherent safety and transmutation capabilities.
- The identification and review of solutions for the improvement of ESFR safety
- The identification and study of innovative options for the main reactor components and fluids.
- The knowledge dissemination through European seminars and doctoral dissertations on the key issues of fast cooled fast reactors: safety, technology and physics.
- The production of around 100 technical reports on the different items covered by the project
Project Context and Objectives:
Objectives and description of the project
The key research goals for fourth generation of European sodium cooled fast reactors can be summarized as follows:
• An improved safety.
The expected progresses of the SFR’s safety architecture have to allow this technology reaching a level at least equivalent, or even higher, to that of the third generation’s reactors. A better consideration within the design of the specificities of sodium cooled reactors with fast neutrons has to be achieved. These specificities concern in particular the risk to meet prompt criticality in severe accident conditions and the risks related to the use of sodium. Potential initiators of abnormal conditions must be reduced to the largest amount reasonably achievable in order to prevent as far as possible incidents and accidents. The severe plant conditions have to be kept highly improbable and the risk of mechanical energy releases during these events and event sequences must be prevented or minimized. The system shall have an increased resistance against the external hazards.
The design shall look for a full in service inspection and reparability capability for the components and systems which are critical from the point of view of the safety. Thoughts on this subject have to consider simultaneously availability and investment protection.
The achievement of a robust architecture vis à vis of abnormal situations and the robustness of the safety demonstrations will materialize the safety improvement.
• A financial risk comparable to that of the other means of energy production.
The economic competitiveness and the reliability of the system has to further progress so that :
o a possible residual additional cost vis à vis of third generation reactors available on the market becomes acceptable for the utilities (considering the “sustainable” character of the technology),
o the investment protection is adequately guaranteed
It is thus necessary to reduce the electricity production cost to a level similar to those of the third generation light water reactors by simplifying the system, reducing the mass of steel and increasing the performances, both increasing the level of energy conversion efficiency and/or increasing the fuel burn-up.
The availability and the reliability of the installation would have to be increased through improved operation and maintenance thanks to an easier and faster inspectability and reparability of components and systems. The life expectancy has to achieve 60 years and dismantling has to be taken into account in the design from the very beginning looking for operation simplification, reduction of waste and of doses to operators.
Finally the design has to look for limiting the risk of a premature “administrative” shut down of the installation through consideration of specific design objectives to address non nuclear accidents (e.g. the risks connected to the sodium) and, generally speaking, an improved acceptability by the public. The acceptance by the public opinion is expected to be strongly connected to the demonstration of the capability to manage waste and to guarantee the durability of the resources (sustainability). Similarly Proliferation Resistance and Physical Protection (PR&PP) concerns have to be addressed to facilitate this acceptance.
Finally it is agreed that simplicity and robustness of the safety demonstrations are essential features to achieve public acceptance.
• A flexible and robust management of the nuclear materials.
The fast neutrons allow using all the uranium available, including the reserve of depleted uranium. To reduce the risks of proliferation the design shall look for systems providing the needed and sufficient plutonium quantity for plant operation, with adequate and possibly innovative fissile materials protection measures.
The management of the nuclear materials allows considering the minimisation of the impact on the environment which has to be as low as practicably achievable. The reduction of the impact on the environment includes on one hand the possibility of transmuting the radioactive long life waste and, on the other hand, progressing in the reduction of the quantity of effluents and the staff radiological exposure. The potential of transmutation with the ESFR of the radioactive waste with long life must be demonstrated.
All these issues were addressed by the project. In addition, it contributed to rebuilt the European expertise on sodium cooled fast reactors
The project was organized in 6 Sub projects:
- SP0: Management (administrative & technical activities)
- SP1: Consistency and assessment
- SP2: Fuel, fuel element, core & fuel cycle
- SP3: Safety concept options and PR & PP issues
- SP4: Innovative Reactor Architecture, components and BOP
- SP5: Education & training
Project Results:
Main results of the project
The main results since the beginning of the project are:
- Detailed definition of starting points for the studies (working horses) for the core and the reactor.
- The identification of core design options to improve inherent safety and transmutation capabilities.
- The identification and review of solutions for the improvement of ESFR safety
- The identification and study of innovative options for the main reactor components and fluids.
- The knowledge dissemination through European seminars and doctoral dissertations on the key issues of fast cooled fast reactors: safety, technology and physics.
- The production of around 100 technical reports on the different items covered by the project.
Core optimization.
Both oxide and carbide cores were defined (Working Horses) as basis for the development on improvements on safety related parameters and minor actinides recycling.
Significant progress has been made in optimising both the oxide and carbide ESFR cores.
The oxide working horse core already had a low reactivity loss over the fuel cycle, which has benefits for rod withdrawal faults, so the optimisation process concentrated on the reduction of sodium void reactivity. The optimisation of the axial configuration has provided a significant overall reduction in sodium void reactivity.
Compared with the Working Horse Carbide Core, the optimized carbide core has undergone significant changes to bring its performance more into line with that of the oxide core. The modifications from the Working Horse Carbide Core have resulted in lower fuel enrichments, nearer to those of the oxide core, with a low reactivity loss over the fuel cycle. The similar performance of the oxide and carbide cores is achieved with a significantly smaller fuel pin diameter and fuel S/A pitch in the carbide core due to the higher density of the carbide fuel. However, at this stage the carbide core has not undergone the same level of optimization as the oxide core.
When comparing safety performance, characterized by sodium void and Doppler constant the carbide core is close to that of the oxide, but not quite as good. CSD absorber rod worth are similar but the DSD array worth is higher in the carbide core than the oxide core. Whilst fuel temperatures have not been assessed, the carbide core can be judged to have lower fuel temperatures during normal operation, because of the greater thermal conductivity of the carbide fuel, giving larger margins to fuel melt.
The carbide working horse core had a significantly higher reactivity loss over the fuel cycle compared with the oxide. By increasing slightly the fuel pin diameter, whilst still retaining the advantages of lower fuel temperatures of carbide fuel, and making some changes to the core layout, the reactivity loss over the cycle has been reduced to a level similar to that of the oxide core. By adopting the CONF-2 axial configuration initially developed for the oxide core, the sodium void reactivity of the carbide core has also been reduced significantly.
Introduction
The development of the ESFR core took place in two phases. In the first phase, lasting two years, technologies for improving core performance, particularly in the areas of safety and minor actinide (MA) transmutation have been studied. The initial studies were based on “working horse” cores which were considered to have reasonable performance and thus would be adequate for the initial work. In the second phase, also lasting two years, the cores were optimised to have improved performance and safety using the work of the first phase.
For the optimisation, two cores were considered: the first using oxide fuel and the second, a more speculative design, using carbide fuel. Because oxide cores have been studied in detail in the past, the oxide working horse core was in a more mature state of development than that of the carbide core. The majority of the studies in the first phase have been based on the oxide core, though the technologies that have been developed are transferable to the carbide core. In this second phase, whilst optimisation work was carried out on the carbide core, the carbide core was not taken to the same level of development as the oxide core.
Criteria for Optimization
A major objective for the optimized core is to improve safety compared with earlier fast reactor concepts. The following criteria have been applied during the optimization process for the core:
1. Compared with the previous generation of fast reactor concepts, a reduction in sodium void reactivity is required to improve performance during un-tripped transient, particularly those involving loss of coolant flow.
2. The amount of reactivity held up by the control rods should be small to limit the impact of accidental control rod withdrawal. Since a major function of the control rods is to take up the excess reactivity of the core, which is greatest at start of cycle, the best means of limiting the amount of reactivity held up in the control rods is to reduce the reactivity loss of the core as the fuel is irradiated. This requires that the core be designed to have low fuel enrichments in the range of 12% to 14%. This leads to the use of fuel pins with large diameters (>10 mm) and a core with a large inventory. This criterion for the control rods is also consistent with an internal breeding gain within the active core of approximately zero; i.e. the reactor is self-sufficient in plutonium without the use of blankets.
3. The core should be capable of transmuting minor actinides. Whilst the inclusion of minor actinides will tend to worsen reactivity coefficients, their introduction should be done in such a way as to limit the impact on safety.
The Oxide Core
The optimisation process started from the oxide working horse core. The core has a power output of 3600 MWth. Axially the core has a fissile height of 1m and contains upper and lower axial blankets.
The fuel sub assembly consists of 271 pins clad in ODS steel with an outer diameter of 10.73 mm and containing an oxide fuel pellet of 9.43 mm outside diameter. The fuel pins are surrounded by a 4.5 mm thick hexagonal EM10 wrapper tube of 206.3 mm outside across flats. The sub-assembly pitching is 210.8 mm.
The oxide working horse core contains large diameter fuel pins with a low fuel enrichment and has a low reactivity loss with burnup over the cycle, meeting already Criterion 2. The optimisation of the core has therefore concentrated on sodium void reactivity reduction.
Fig. 1. Oxide working horse core – Radial cross section
Fig. 2. Oxide working horse core – Axial cross section
Sodium Void Reduction
In the first phase of the study, various means of reducing sodium void reactivity were studied. The study considered radial and axial heterogeneity, the use of moderator to soften the neutron spectrum, reduction in the height to diameter ratio of the core to increase neutron leakage, and the introduction of an upper sodium plenum with an absorber layer above the plenum to absorb neutrons when the plenum has voided. The initial conclusion from the study were that the upper sodium plenum with the absorber showed the most promise for sodium void reactivity reduction without detrimental effects on other areas of performance. SIMMER coupled neutronics and thermal-hydraulics analysis was then undertaken, also in for the core with sodium plenum to confirm that in loss of flow events, the voiding of the plenum limited the reactivity of sodium voiding and ultimately contributed to the shutdown of the core. As a result of the work a recommendation was made for the inclusion of an upper sodium plenum of 600 mm with a 300mm thick absorber layer above the plenum, called the CONF-2 axial configuration.
Some additional work on sodium void reduction by core geometry change, has been undertaken by EDF. The work used an RZ geometry model and considered stepped height core with a reduction in active core height to 0.85 m in the inner core whilst increasing the height to 1.15m in the outer core to retain the total core volume. This increases the effectiveness of axial leakage of neutrons from the inner core and increases the surface area from which neutrons can leak from the outer core. The analysis demonstrated a possible further reduction in sodium void reactivity of 0.6$ to 0.8$ but concluded that more detailed 3D analysis and transient studies would be required to confirm the effectiveness of such a sodium void reduction measure.
Minor Actinide Transmutation
Studies of a wide range of both homogeneous and heterogeneous options for minor actinide transmutation were undertaken in the first phase of the project. The options covered a wide range of transmutation rates. For the optimised core a choice has been made of two options; the first for a low minor actinide transmutation rate and the second a high transmutation rate. These options, from Ref 4, are known as the HET2 configuration and the HOM4 configuration.
As a minimum position for the recycling of minor actinides, it would be expected that the ESFR core should be capable of transmuting its own minor actinides. The configuration appropriate to this level of consumption is the HET2 configuration. The HET2 configuration consists of a ring of 84 radial blanket sub-assemblies.
The radial blankets contain fuel composed of 20% MA and 80% depleted uranium. This configuration, with the MA target sub-assemblies outside the active core will have a small impact on the reactivity coefficients of the core.
A high level of minor actinide consumption is provided by the HOM4 configuration. This is capable of transmuting not only the minor actinides produced within the ESFR core but also minor actinides from thermal reactors. The HOM4 configuration consists of 4% by weight of MA homogeneously included within the core fuel, and will have a greater impact on reactivity coefficients than the HET2 configuration.
By choosing the HET2 and HOM4 configurations from the options studied in Ref 4 for the optimised core a wide range of MA transmutation will be considered. These two options covers the full range of impact on the safety performance of including MA transmutation in the reactor and allow judgements to be made as to the safety consequences of MA transmutation.
More detailed analysis of the minor actinide transmutation capabilities of the oxide core for the HET2 and HOM4 configurations, using the CONF2 axial configuration, has been undertaken in the second phase of the work. The impact of MA loading on safety, particularly sodium void reactivity and Doppler constant, in addition to the cores’ efficiency in transmutation of MA has been considered. In addition to the standard configuration of HOM4, which did not include MA in the lower axial blanket, the Ciemat analysis also considered loading of 4% MA into the lower axial blanket of the HOM4 case to further improve MA consumption.
The conclusions to be drawn from this analysis are as follows:
- The HET2 configuration consumes approximately the MA produced by the oxide core. That is the core is approximately neutral in MA production. When this is looked at in more detail there is a small net consumption of Neptunium and Americium and a small net production of Curium.
- The impact on safety of MA in the HET2 configuration, as judged by the sodium void reactivity and Doppler constant, is minimal.
- As expected, the HOM4 configuration has a much higher MA consumption. Over 2050 EFPD, the life of the fuel, the MA transmuted amounts to 870 kg within the active core. Including 4% MA in the lower axial blanket transmutes a further 175 kg. There is, however, a net Curium production in the core and lower blanket of approximately 214 kg.
- Homogeneous loading of the does have a detrimental effect on sodium void reactivity and Doppler constant: sodium void reactivity is increased by approximately 100 pcm and the magnitude of the Doppler constant is reduced by approximately 140 pcm. It should be noted that with the reduction in total sodium void reactivity by the inclusion of the upper plenum of the CONF2 axial configuration and the low enrichment of the fuel in this core, the absolute values of the sodium void reactivity and Doppler constant with homogeneous minor actinide content may still be acceptable. Transient analysis would be required before a final judgment could be made.
Rod Worths
An assessment of the control rod worth of the ESFR oxide core has been undertaken (See Appendix D for details) to ensure that shutdown can be achieved. Assessments of control rod worth for the primary (CSD) and secondary (DSD) CRSs were performed using MCNP and include the determination of the worth curves as a function of rod bank insertion. The assessments were performed for Beginning of Equilibrium Cycle (BOC) condition.
Allowing for a single stuck rod in either the CSD or DSD arrays, the calculated worth of the arrays is as follows:
CSD : 4535 pcm
DSD : 1260 pcm
The requirement to shut down to cold temperatures (100 °C) was calculated and is 2084 pcm.
The CSDs also have to be capable of covering the reactivity loss with burnup, this is approximately 200 pcm, even for a single batch refueling scheme. It is clear that the total requirement of ~2300 pcm can be accommodated by the CSDs with a significant margin when allowing for uncertainty on the calculated values. The DSDs do not need to take up the reactivity loss with burnup since that is done by the CSDs. They do need to cover the reactivity requirement on shutting down to lower temperatures.
Conventionally, the requirement for the CSDs has been to shut down to sodium inlet temperature rather than cold, which would give a requirement of rather less than 2000 pcm. With the array of 9 DSD rods used at present this is not possible. It is judged likely that some transfer of rods from the CSD group to the DSD group would permit the DSDs to meet their requirements without affecting the ability of the CSDs to meet theirs, because of the significant margins available in the current CSD array.
Cores Containing Features for Molten Fuel Discharge
The most efficient options to reduce the void effect, without strongly influencing other parameters, were considered. The studies led to the so called CONF2 configuration. In particular, the axial core structure was modified by the introduction a lower fertile blanket, a larger Na plenum shifted closer to the core, and an absorber layer above the plenum.
A preliminary study of further core design options has been undertaken, related to incorporation of special elements (SAs) with the main aim of facilitating the discharge of molten fuel after a hypothetical severe accident.
The starting point for the study was the CONF2 configuration without MAs. The study provides an evaluation of the neutronics characteristics of these configurations, mainly at BOL conditions, in particular sodium void reactivity and power profiles. The other parameters (Doppler and kinetics parameters, βeff and Λ) are expected to remain practically unchanged.
In particular the following two configurations have been considered:
- CONF3B: where 18 special SAs (wrapper with only Na inside) are introduced in the inner core. 18 fuel SAs have been added to the outer core periphery in order to maintain the total number of fuel SAs.
- CONF3C: the same configuration as CONF-3B, but adopting different special SAs: with the same design as fuel SAs (271 pins, 1.173 cm pitch) but with pins filled with helium instead of fuel.
In both cases, the fuel enrichment for the computation models was the same as in CONF2, as k-eff is not strongly affected by these modifications.
The study has confirmed a high potential of the ESFR core design and, in particular of CONF2, to be a basis for further optimizations. The modifications reduced the core and extended (core and plenum) void effects in CONF3B and CONF3C by about 100 pcm compared with CONF2. If voiding of the special SAs occurs in the case of molten fuel ingress, this also leads to a reactivity reduction. A benchmark against MCNP has shown that the MCNP results are more conservative than the ERANOS ones, except for voiding special SAs. A higher void effect reduction is possible due to incorporation of a larger number of special SAs.
The adoption of special SAs leads to minor power profile variations at BOL. These variations can be reduced by increasing slightly the inner core enrichment. The Power profile at EOEC remains almost the same.
Results of transient analyses show that incorporation of special SAs may improve the core safety performance. Larger effort is needed to get a better understanding on the consequences of the introduction of special SAs under severe accident conditions.
The Carbide Core
The core has a power output of 3600 MWth. Axially the core has a fissile height of 0.8m and contains upper and lower axial blankets.
The fuel sub assembly consists of 331 pins clad in ODS steel with an outer diameter of 8.0 mm and containing a carbide fuel pellet of 6.87 mm outside diameter. The fuel pins are surrounded by a 4.5 mm thick hexagonal EM10 wrapper tube of 178.7 mm outside across flats. The sub-assembly pitching is 178.7 mm.
The performance of the carbide Working Horse core is rather different from that of the oxide core: it has a significantly higher reactivity loss with burnup and a much lower internal breeding gain.
Modifications to produce a more optimised performance of the carbide core are discussed in the following sub-section.
Initial Optimisation of the Carbide Core
The carbide working horse core is a less mature design than the oxide core and required further optimisation. The limits on fuel residence time were defined in the working horse core by the maximum clad damage permitted by ODS, which is 200 dpa. This leads to a peak fuel burnup in excess of 20%, which is rather optimistic for a type of fuel for which there is much less irradiation experience than for oxide. This should be considered to be an interim position until more information becomes available on carbide fuel performance.
The particular criteria for optimisation of the carbide core were as follows:
1. The reactivity loss between start and end of cycle should be reduced to a low level to limit the amount of reactivity that can be injected into the core on accidental withdrawal of one or more control rods.
2. The internal breeding gain of the core should be increased to a level similar to that of the oxide core (~0.02). This is consistent with Criterion 1.
3. Reduce the peak burnup of the fuel to ~15%, as already discussed.
4. Retain the safety advantages of the carbide core which arise from the high thermal conductivity of the fuel, i.e. normal operational fuel temperatures should be lower than those of the oxide core. This can be achieved by ensuring that peak linear ratings do not exceed those of the oxide core.
To reduce the large reactivity loss with burnup, the optimised core height was increased from 0.8m to 1.0m (like that of the oxide core) and the fuel pin outer diameter was increased to 8.5 mm. This has allowed the fuel enrichment to be reduced significantly so that the reactivity loss between start and end of equilibrium cycle when using a 3-batch fuel cycle is only 440 pcm.
The reduction of sodium void reactivity, studied particularly for the oxide core, would be expected to be equally applicable to the carbide core. The CONF-2 axial configuration has therefore been adopted for the carbide core.
The advantages to using carbide fuel arise from its higher density and greater thermal conductivity than oxide. The higher fuel pellet density permits low enrichment to be attained with a smaller fuel pin diameter than the oxide core. The higher thermal conductivity provides greater margin to fuel melting temperature in fault transients, provided that normal operation linear ratings are maintained at a similar level to those of the oxide core. Both of these advantages have been retained in the optimised core design.
The specification of the optimised core is given in Table 1, where the design parameters are compared with those of the carbide Working Horse core. The optimised core plan is given in Figure 3. Because the subassembly pitch has been reduced slightly, minor modifications have been made to the CSD and DSD rods.
The revised dimensions have been determined by scaling of the carbide Working Horse core control rod dimensions.
Rod Worths
An initial assessment of the control rod worth of the ESFR carbide core has been undertaken to ensure that shutdown can be achieved. Assessments of control rod worth for the primary (CSD) and secondary (DSD) CRSs were performed using MCNP and include the determination of the worth curves as a function of rod bank insertion. The assessments were performed for Beginning of Equilibrium Cycle (BOC) condition.
Allowing for a single stuck rod in either the CSD or DSD arrays, the calculated worth of the arrays is as follows:
CSD : 4875 pcm
DSD : 2046 pcm
The requirement to shut down to cold temperatures (100 °C) was calculated and is 1585 pcm.
The CSDs also have to be capable of covering the reactivity loss with burnup. This is approximately 650 pcm, even for a single batch refueling scheme. It is clear that the total requirement of ~2300 pcm can be accommodated by the CSDs with a significant margin even when allowing for uncertainty on the calculated values. The DSDs do not need to take up the reactivity loss with burnup since that is done by the CSDs. They do need to cover the reactivity requirement on shutting down to lower temperatures.
Conventionally, the requirement for the CSDs has been to shut down to sodium inlet temperature rather than cold, which would give a requirement of rather less than 1600 pcm. With the array of 9 DSD rods used at present this is possible, allowing for uncertainties on the calculated values.
Comparison between the Oxide and Carbide Cores
Compared with the Working Horse Carbide Core, the optimized carbide core has undergone significant changes to bring its performance more into line with that of the oxide core. The modifications from the Working Horse Carbide Core have resulted in lower fuel enrichments, nearer to those of the oxide core, with a low reactivity loss over the fuel cycle. The similar performance of the oxide and carbide cores is achieved with a significantly smaller fuel pin diameter and fuel S/A pitch in the carbide core due to the higher density of the carbide fuel. However, at this stage the carbide core has not undergone the same level of optimization as the oxide core.
When comparing safety performance, characterized by sodium void and Doppler constant the carbide core is close to that of the oxide, but not quite as good. CSD absorber rod worths are similar but the DSD array worth is higher in the carbide core than the oxide core.
Whilst fuel temperatures have not been assessed, the carbide core can be judged to have lower fuel temperatures during normal operation, because of the greater thermal conductivity of the carbide fuel, giving larger margins to fuel melt.
Conclusions core optimization activities
Significant progress has been made in optimising both the oxide and carbide ESFR cores.
The oxide working horse core already had a low reactivity loss over the fuel cycle, which has benefits for rod withdrawal faults, so the optimisation process concentrated on the reduction of sodium void reactivity. The CONF-2 axial configuration has provided a significant overall reduction in sodium void reactivity. The working horse oxide core with the CONF-2 axial configuration been adopted as the optimised oxide core.
The carbide working horse core had a significantly higher reactivity loss over the fuel cycle compared with the oxide. By increasing slightly the fuel pin diameter, whilst still retaining the advantages of lower fuel temperatures of carbide fuel, and making some changes to the core layout, the reactivity loss over the cycle has been reduced to a level similar to that of the oxide core. By adopting the CONF-2 axial configuration initially developed for the oxide core, the sodium void reactivity of the carbide core has also been reduced significantly.
Fuel
Nitrides/Carbides fuels
Tests were conducted on the production of (U,Pu,Am)X fuels (X = C and N). the sol gel route was used to produce precursor (U,Pu)O2 beads with carbon in place for the carbothermal reduction. The Pu content Pu/(U+Pu) content was about 20%. Two batches were available with C/(U+Pu) molar ratios in excess of 2.0 and 3.0 for the synthesis of the carbides and nitrides respectively. Both sets of material were infiltrated with an Am nitrate solution, with the beads then being thermally treated under Ar/H2 to convert the Am nitrate to oxide. At this stage their composition can be described as a mixture of (U,Pu)O2, AmO2 and Carbon. Given that the risk of carbon volatisation exists, a slightly higher Am content Am/(U+Pu+Am) than to be found in typical fuels for the homogeneous recycling of minor actinides (i.e. 5%) was chosen. Thus the Am content was of the order of 8%. During the carbothermal synthesis it was expected that this value would diminish slightly, especially for the carbide fuel.
The samples were then removed to a dedicated production facility, where the glovebox atmosphere can reach the required O2 and H2O impurity levels (typically <10 ppm). Carbothermal reduction tests were performed on the (U,Pu) precursors themselves. Good quality carbide and nitride samples have been successfully produced, though a small amount of oxide impurity could always be detected. The corresponding (U,Pu,Am)X samples were also synthesised, and at the time of writing this report are being characterised.
Am-bearing Oxide fuels.
Sol-gel infiltration process and physical property measurements:
The production of the (U,Am)O2 samples was completed earlier in the project using a method based on the infiltration of UO2 beads, with subsequent calsination, compaction and sintering steps. The thermal conductivity of the samples has also been achieved using the JRC- ITU laser flash (LAF) method.
Powder metallurgy process and physical property measurements:
The experimental grid includes four AmBB compositions with two contents of Am (15 and 20%), two O/M ratios (for the 20% Am composition) and two densities (for the 15% Am composition). The two densities correspond to a standard dense microstructure and an optimized one with a low density and a significant part of opened porosity in order to favour gaseous release under irradiation. Three compositions were fabricated and characterized at the ATALANTE facility (CEA Marcoule) during the considered period. It is recalled that the composition with 15% of Americium and the optimized microstructure was fabricated and characterized during the previous period. During this last period, difficulties were encountered when fabricating dense AmBB material. A new method, patented as the UMACS process, was developed successfully for this purpose during the considered period and allowed the fabrication of dense AmBB samples for the CP-ESFR programme. Characterization measurements on fabricated samples include chemical analysis, density measurements and XRD analysis. Samples from the four compositions were sent to the LEFCA laboratory at the CEA Cadarache for physical property measurements. Thermal conductivity and thermal expansion measurements are currently on-going at the LEFCA laboratory. Some samples issued from the composition n°3 (15% Am – dense material) were also sent to the JRC-ITU for melting point and vapour pressure measurements.
Nitride and carbide MA fuel inter-comparison
The in-core behaviour of carbides and nitrides has been assessed based on a literature review of out-of-pile accessible irradiation experiments.
SAFETY
The activities on safety was divided in six areas
Safety objectives and design principles
Technology neutral safety objectives and principles consistent with the European safety framework for new nuclear plants (e.g. EUR, Technical Guidelines considered for EPRTM) were provided. The approach is deterministic and complemented by probabilistic studies. Sodium cooled Fast Reactor (SFR) technology operational and licensing feedback has to be considered and specific risks (in particular, those linked with sodium use) have to be addressed. Basic rules for deterministic safety analyses of internal events, consideration of hazards, safety classification and qualification of components were proposed. The role of Probabilistic Safety Assessments is defined. The design strategy with respect to safety was also adressed: guidelines for the implementation of the main safety functions (reactivity control, decay heat removal, confinement) were provided, then the approach for identification of challenging events (i.e. with a potential of large radiological releases) were applied and relevant possible design strategies proposed with respect to the consideration of whole core accidents.
Defense-in-depth levels and incidents/accidents for Design and Beyond Design Basis events
This activity was focused on the safety goals and general safety principles for future reactors, especially applicable for the ESFR and the characteristics to achieve a robust safety architecture.
For the design of a future reactor, a prerequisite is related to how hypothetical situations of general core meltdown are dealt with. The way to deal these situations has namely to be considered early, because of its prime importance on the design options. So this issue was discussed presenting two possible strategies on this item. The main safety issues related to SFR were analysed on the basis of operating and licensing feedback.
Studies of representative transients and accident scenarios for DB and BDB events
A specific task was devoted to the development multi-physics models for the plant design and the calculation of some postulated plant transients. The task group consisted of eight organisations namely, CEA (France), ENEA (Italy), EDF (France), UPM (Spain), PSI (Switzerland), KIT (Germany), NRG (Netherlands) and JRC-IET (European Commission). Each organisation provided its expertise on the use of system codes for sodium-cooled fast reactor applications. The computational tools, originally developed for commercial light water nuclear reactor safety analyses, have undergone several modifications (coolant material properties, heat transfer and pressure drop model) to be applied for SFR studies. The approach followed in the task was in the first instance to set up a benchmarking exercise where each organisation amended its own system code models of the plant and then compared its calculation results with those of the other partners on a commonly agreed basis. In this respect, this benchmark exercise provided the unique opportunity to compare the performance of the current state-of-the-art system codes for applications related to sodium-cooled fast reactor safety analyses.
The benchmarking activity was followed by the calculation of eight significant transients:
1. Control rod withdrawal
2. Primary pumps acceleration from 30% nominal conditions
3. All secondary pumps trip
4. Total loss of steam generators feedwater
5. Flow bypass of the core
6. Loss of offsite power
7. Partial S/A blockage
8. Pump/core feed duct rupture
Transients for the design extension conditions (DEC) were also investigated
1. Unprotected loss of flow
2. Instantaneous total S/A blockage.
Provisions to mitigate core degradation
A review of sacrificial material candidates has been conducted, considering both neutron absorber and diluent materials. This review was initially based on criteria related to thermophysical and chemical thermodynamic properties (melting and boiling temperature, ability to form a liquid solution with molten fuel…). Neutronic calculations have been done for some generic configurations in order to estimate the reactivity decrease due to the mixing of corium with some different materials. Materials such as aluminium oxide, uranium oxide and hafnium or europium oxides were presented in more details and their relative advantages discussed.
Mass (t) of Eu2O3 needed to avoid re-criticality as a function of non-melted core fraction (%)
Containment and core catcher
The in vessel design of historical and planned SFRs was reviewed with respect to corium relocation paths from the core to the core catcher. The corium source term was assessed based on recent core designs and on preliminary results of SAS-4A calculations, since presently not quantitative information are available on corium masses ejected upwards and downwards from the core. It turned out, that due to the selected core configuration, the ejected corium mass out of the S/A region into the hot plenum is negligible in that sense, but it can be of importance in case of agglomeration of corium. The coolability is strongly dependent on the geometry of the trap and integrity and geometry of the in-vessel structures. Downward relocation of corium was discussed only briefly, depending on the presently available data of the geometry in diagrid and strongback.
The work was based on available data for the core configurations of CP-ESFR, and available results of SIMMER calculations were also considered.
Different mitigation approaches to face a core disruption accident (CDA)
Modelling capabilities
The activity aimed at evaluation of the modeling capabilities of accident scenarios in the frame of the ESFR defence in depth approach using the PIRT (Phenomena Identification Ranking Tables) methodology. The identification of the phenomena that could affect:
- core degradation and in-vessel structure behavior during transient of the design basis domain, potentially leading to local extended core melting accidents;
- releases and transfer of radionuclides under hypothesized accident conditions;
- sodium fire occurrence and evolution;
- sodium chemical reaction occurrence and evolution (with water, air and concrete), was performed as well as evaluation of their knowledge status based on the basis of the past RD work,
The intent was clearly to identify additional research needed for licensing and risk assessment purposes. The importance of each phenomenon was qualitatively evaluated and ranked from a safety point of view as well as the need for additional experimental research. In all, 25 phenomena that are of high importance and have a high need for additional experimental research were identified.
Innovative reactor architectures, components and balance of plant
Plant concept and layouts
Two reactor concepts (one pool and one loop type) where defined and described in detail and then used as starting points for the studies on innovations. Their advantages and drawbacks where compared in the European context. In conclusion, the loop concept would be probably in a lower power range compared to the pool concept due to the high motivation to limit the number of loops and therefore to have large size components. Potential advantages of the loop type (earthquake response, manufacturability in plants, free sodium level constraints) needs to be balanced by the drawbacks linked to gas entrainment and fuel handling issues. Clearly a large investment on loop type issues would be necessary to reach the same level of maturity reached from the much larger European experience on pool type reactors.
Specific studies were devoted to innovative components like the inner vessel, the isolation condensers allowing the improvement the availability and safety by the diversification of decay heat removal systems.
Several alternative designs for the components of the heat transfer system were analysed like kidney shape intermediate heat exchangers allowing a vessel diameter reduction. Integrated pump/exchanger components were considered for cost reduction. Other technologies like electromagnetic pumps and inverted steam generators (sodium circulating inside the tubes) were studied.
Electromagnetic pump with removable outer inductor
Energy conversion systems
Both cycles, water/steam and gas where considered.
The work was focused on the review of incentives and pre-design of Steam Generators minimizing the risks of large sodium water reaction. Considered solutions were double-wall SG with/without leak detection, Inverted type SG with sodium inside tubes, double tube with solid matrix. Concerning inverted SG, a feedback from past experience has been provided as well as design optimization and extrapolation to larger power units.
The evaluation of double component (IHX/SG) technology with coupling fluid was performed for which studies included a comparison between design candidates, a review of the safety and operational implications including corrosion and erosion, risks of plugging and leak detection issues. Methods for SG leak detection as well as ISI including innovative SG were reviewed.
Double component featuring IHX and SG bundles couple by circulating Pb-Bi
Crosscut activities - Tools and experiments in support to the design and codes and standards.
The activity was focused on three areas:
- Review of tools and experiments needs in support of new safety related systems and components
- Review of tools and experiments needs on thermal-hydraulics, mechanics, chemistry
- Update of Codes & Standards in support of ESFR needs including: State of the Art of existing Codes & Standards
R&D needs for primary inner vessel, cover gas and circuits have been evaluated. For the energy conversion system the different steam generator design have been summarised and an R&D map was prepared. Finally, innovative solutions proposed by the ESFR partners for the decay heat removal system, the external vessel storage tank of the fuel handling system and the core catcher have been assessed. The identified R&D roadmap was summarised together with the available facilities and tools.
Links with the European ADRIANA project have been identified: the list of needs to be addressed by R&D, in support to the development of SFR (and more particularly to French ASTRID project), the list of existing facilities, dedicated to SFR development (available or to be refurbished), the list of facilities, to be designed and/or implemented. All these facilities have been ranked, and a road-map has been established. A list of European facilities, criteria for ranking, priorities for investment, first cost evaluations have also been established
Current state of the art in terms of Codes and Standards (RCC-MR, ASME, EN13445, etc) and have identified actions required on the R&D side and on the codification side to respond to specific needs either resulting from new design options or from evolution of regulatory framework (European Pressure Equipment Directive and associated national regulation).
Education and training
Workshops
The following workshops were organized:
- CEA - ESFR Na and other coolants: in-reactor behaviour and safety
- CEA . ESFR SFR: functional analysis and design safety
- KIT Uni-Karlsruhe - Education and Training Workshop on Model Verification & Validation Principles & Application to ESFR
- KIT Uni-Karlsruhe - Education and Training Workshop on Model Calibration & Predictive Estimation Principles & Application to ESFR (these two last workshops were merged in only one)
- UNI Rome - Workshop on ESFR Engineering Aspects
- Universitad Politécnica de Madrid - Workshop on ESFR Reactor Physics and Safety project results.
More than 150 European young engineers and students attended these successful workshops.
First CP ESFR Workshop at CEA Cadarache
Doctoral dissertations
Five doctoral dissertations were carried out:
- Uni-Karlsruhe
Verification and Experimental Validation of ESFR-core thermal-hydraulics codes and models
- NRG
Sodium fire assessment.
- CEA
Sodium-water interaction, in presence of inert gas or not.
- CEA
Experimental and validation of SFR severe accident phenomenology (finished)
- UPM
Development and verification of European (NURESIM-NURISP) pin-by-pin and nodal coupled NK-TH simulation codes and integrated platform with demonstration applications for ESFR core physics and safety analysis
Potential Impact:
As the end-products developed by the project may be used not only for R&D activities but also for industrial applications, many European industry and safety authorities (or technical safety organizations) are contributing to CP ESFR.
Progressively the research activities on sodium cooled fast reactors at European level are strongly coordinated by the project, which will contribute to an optimised use of European resources.
Through education and training programmes, CP ESFR is developing synergies with educational institutions and thus keeping attractive the concerned domain of activity for students and young researchers. This will also contribute to enhance and preserve in a sustainable way the European scientific leadership.
List of Websites:
http://www.cp-esfr.eu/
The key research goals for fourth generation of European sodium cooled fast reactors can be summarized as follows:
• An improved safety.
• A financial risk comparable to that of the other means of energy production.
• A flexible and robust management of the nuclear materials.
All these issues were addressed by the project. In addition, it contributed to rebuilt the European expertise on sodium cooled fast reactors
The project was organized in 6 Sub projects:
- SP0: Management (administrative & technical activities)
- SP1: Consistency and assessment
- SP2: Fuel, fuel element, core & fuel cycle
- SP3: Safety concept options and PR & PP issues
- SP4: Innovative Reactor Architecture, components and BOP
- SP5: Education & training
Main results of the project
The main results since the beginning of the project are:
- Detailed definition of starting points for the studies (working horses) for the core and the reactor.
- The identification of core design options to improve inherent safety and transmutation capabilities.
- The identification and review of solutions for the improvement of ESFR safety
- The identification and study of innovative options for the main reactor components and fluids.
- The knowledge dissemination through European seminars and doctoral dissertations on the key issues of fast cooled fast reactors: safety, technology and physics.
- The production of around 100 technical reports on the different items covered by the project
Project Context and Objectives:
Objectives and description of the project
The key research goals for fourth generation of European sodium cooled fast reactors can be summarized as follows:
• An improved safety.
The expected progresses of the SFR’s safety architecture have to allow this technology reaching a level at least equivalent, or even higher, to that of the third generation’s reactors. A better consideration within the design of the specificities of sodium cooled reactors with fast neutrons has to be achieved. These specificities concern in particular the risk to meet prompt criticality in severe accident conditions and the risks related to the use of sodium. Potential initiators of abnormal conditions must be reduced to the largest amount reasonably achievable in order to prevent as far as possible incidents and accidents. The severe plant conditions have to be kept highly improbable and the risk of mechanical energy releases during these events and event sequences must be prevented or minimized. The system shall have an increased resistance against the external hazards.
The design shall look for a full in service inspection and reparability capability for the components and systems which are critical from the point of view of the safety. Thoughts on this subject have to consider simultaneously availability and investment protection.
The achievement of a robust architecture vis à vis of abnormal situations and the robustness of the safety demonstrations will materialize the safety improvement.
• A financial risk comparable to that of the other means of energy production.
The economic competitiveness and the reliability of the system has to further progress so that :
o a possible residual additional cost vis à vis of third generation reactors available on the market becomes acceptable for the utilities (considering the “sustainable” character of the technology),
o the investment protection is adequately guaranteed
It is thus necessary to reduce the electricity production cost to a level similar to those of the third generation light water reactors by simplifying the system, reducing the mass of steel and increasing the performances, both increasing the level of energy conversion efficiency and/or increasing the fuel burn-up.
The availability and the reliability of the installation would have to be increased through improved operation and maintenance thanks to an easier and faster inspectability and reparability of components and systems. The life expectancy has to achieve 60 years and dismantling has to be taken into account in the design from the very beginning looking for operation simplification, reduction of waste and of doses to operators.
Finally the design has to look for limiting the risk of a premature “administrative” shut down of the installation through consideration of specific design objectives to address non nuclear accidents (e.g. the risks connected to the sodium) and, generally speaking, an improved acceptability by the public. The acceptance by the public opinion is expected to be strongly connected to the demonstration of the capability to manage waste and to guarantee the durability of the resources (sustainability). Similarly Proliferation Resistance and Physical Protection (PR&PP) concerns have to be addressed to facilitate this acceptance.
Finally it is agreed that simplicity and robustness of the safety demonstrations are essential features to achieve public acceptance.
• A flexible and robust management of the nuclear materials.
The fast neutrons allow using all the uranium available, including the reserve of depleted uranium. To reduce the risks of proliferation the design shall look for systems providing the needed and sufficient plutonium quantity for plant operation, with adequate and possibly innovative fissile materials protection measures.
The management of the nuclear materials allows considering the minimisation of the impact on the environment which has to be as low as practicably achievable. The reduction of the impact on the environment includes on one hand the possibility of transmuting the radioactive long life waste and, on the other hand, progressing in the reduction of the quantity of effluents and the staff radiological exposure. The potential of transmutation with the ESFR of the radioactive waste with long life must be demonstrated.
All these issues were addressed by the project. In addition, it contributed to rebuilt the European expertise on sodium cooled fast reactors
The project was organized in 6 Sub projects:
- SP0: Management (administrative & technical activities)
- SP1: Consistency and assessment
- SP2: Fuel, fuel element, core & fuel cycle
- SP3: Safety concept options and PR & PP issues
- SP4: Innovative Reactor Architecture, components and BOP
- SP5: Education & training
Project Results:
Main results of the project
The main results since the beginning of the project are:
- Detailed definition of starting points for the studies (working horses) for the core and the reactor.
- The identification of core design options to improve inherent safety and transmutation capabilities.
- The identification and review of solutions for the improvement of ESFR safety
- The identification and study of innovative options for the main reactor components and fluids.
- The knowledge dissemination through European seminars and doctoral dissertations on the key issues of fast cooled fast reactors: safety, technology and physics.
- The production of around 100 technical reports on the different items covered by the project.
Core optimization.
Both oxide and carbide cores were defined (Working Horses) as basis for the development on improvements on safety related parameters and minor actinides recycling.
Significant progress has been made in optimising both the oxide and carbide ESFR cores.
The oxide working horse core already had a low reactivity loss over the fuel cycle, which has benefits for rod withdrawal faults, so the optimisation process concentrated on the reduction of sodium void reactivity. The optimisation of the axial configuration has provided a significant overall reduction in sodium void reactivity.
Compared with the Working Horse Carbide Core, the optimized carbide core has undergone significant changes to bring its performance more into line with that of the oxide core. The modifications from the Working Horse Carbide Core have resulted in lower fuel enrichments, nearer to those of the oxide core, with a low reactivity loss over the fuel cycle. The similar performance of the oxide and carbide cores is achieved with a significantly smaller fuel pin diameter and fuel S/A pitch in the carbide core due to the higher density of the carbide fuel. However, at this stage the carbide core has not undergone the same level of optimization as the oxide core.
When comparing safety performance, characterized by sodium void and Doppler constant the carbide core is close to that of the oxide, but not quite as good. CSD absorber rod worth are similar but the DSD array worth is higher in the carbide core than the oxide core. Whilst fuel temperatures have not been assessed, the carbide core can be judged to have lower fuel temperatures during normal operation, because of the greater thermal conductivity of the carbide fuel, giving larger margins to fuel melt.
The carbide working horse core had a significantly higher reactivity loss over the fuel cycle compared with the oxide. By increasing slightly the fuel pin diameter, whilst still retaining the advantages of lower fuel temperatures of carbide fuel, and making some changes to the core layout, the reactivity loss over the cycle has been reduced to a level similar to that of the oxide core. By adopting the CONF-2 axial configuration initially developed for the oxide core, the sodium void reactivity of the carbide core has also been reduced significantly.
Introduction
The development of the ESFR core took place in two phases. In the first phase, lasting two years, technologies for improving core performance, particularly in the areas of safety and minor actinide (MA) transmutation have been studied. The initial studies were based on “working horse” cores which were considered to have reasonable performance and thus would be adequate for the initial work. In the second phase, also lasting two years, the cores were optimised to have improved performance and safety using the work of the first phase.
For the optimisation, two cores were considered: the first using oxide fuel and the second, a more speculative design, using carbide fuel. Because oxide cores have been studied in detail in the past, the oxide working horse core was in a more mature state of development than that of the carbide core. The majority of the studies in the first phase have been based on the oxide core, though the technologies that have been developed are transferable to the carbide core. In this second phase, whilst optimisation work was carried out on the carbide core, the carbide core was not taken to the same level of development as the oxide core.
Criteria for Optimization
A major objective for the optimized core is to improve safety compared with earlier fast reactor concepts. The following criteria have been applied during the optimization process for the core:
1. Compared with the previous generation of fast reactor concepts, a reduction in sodium void reactivity is required to improve performance during un-tripped transient, particularly those involving loss of coolant flow.
2. The amount of reactivity held up by the control rods should be small to limit the impact of accidental control rod withdrawal. Since a major function of the control rods is to take up the excess reactivity of the core, which is greatest at start of cycle, the best means of limiting the amount of reactivity held up in the control rods is to reduce the reactivity loss of the core as the fuel is irradiated. This requires that the core be designed to have low fuel enrichments in the range of 12% to 14%. This leads to the use of fuel pins with large diameters (>10 mm) and a core with a large inventory. This criterion for the control rods is also consistent with an internal breeding gain within the active core of approximately zero; i.e. the reactor is self-sufficient in plutonium without the use of blankets.
3. The core should be capable of transmuting minor actinides. Whilst the inclusion of minor actinides will tend to worsen reactivity coefficients, their introduction should be done in such a way as to limit the impact on safety.
The Oxide Core
The optimisation process started from the oxide working horse core. The core has a power output of 3600 MWth. Axially the core has a fissile height of 1m and contains upper and lower axial blankets.
The fuel sub assembly consists of 271 pins clad in ODS steel with an outer diameter of 10.73 mm and containing an oxide fuel pellet of 9.43 mm outside diameter. The fuel pins are surrounded by a 4.5 mm thick hexagonal EM10 wrapper tube of 206.3 mm outside across flats. The sub-assembly pitching is 210.8 mm.
The oxide working horse core contains large diameter fuel pins with a low fuel enrichment and has a low reactivity loss with burnup over the cycle, meeting already Criterion 2. The optimisation of the core has therefore concentrated on sodium void reactivity reduction.
Fig. 1. Oxide working horse core – Radial cross section
Fig. 2. Oxide working horse core – Axial cross section
Sodium Void Reduction
In the first phase of the study, various means of reducing sodium void reactivity were studied. The study considered radial and axial heterogeneity, the use of moderator to soften the neutron spectrum, reduction in the height to diameter ratio of the core to increase neutron leakage, and the introduction of an upper sodium plenum with an absorber layer above the plenum to absorb neutrons when the plenum has voided. The initial conclusion from the study were that the upper sodium plenum with the absorber showed the most promise for sodium void reactivity reduction without detrimental effects on other areas of performance. SIMMER coupled neutronics and thermal-hydraulics analysis was then undertaken, also in for the core with sodium plenum to confirm that in loss of flow events, the voiding of the plenum limited the reactivity of sodium voiding and ultimately contributed to the shutdown of the core. As a result of the work a recommendation was made for the inclusion of an upper sodium plenum of 600 mm with a 300mm thick absorber layer above the plenum, called the CONF-2 axial configuration.
Some additional work on sodium void reduction by core geometry change, has been undertaken by EDF. The work used an RZ geometry model and considered stepped height core with a reduction in active core height to 0.85 m in the inner core whilst increasing the height to 1.15m in the outer core to retain the total core volume. This increases the effectiveness of axial leakage of neutrons from the inner core and increases the surface area from which neutrons can leak from the outer core. The analysis demonstrated a possible further reduction in sodium void reactivity of 0.6$ to 0.8$ but concluded that more detailed 3D analysis and transient studies would be required to confirm the effectiveness of such a sodium void reduction measure.
Minor Actinide Transmutation
Studies of a wide range of both homogeneous and heterogeneous options for minor actinide transmutation were undertaken in the first phase of the project. The options covered a wide range of transmutation rates. For the optimised core a choice has been made of two options; the first for a low minor actinide transmutation rate and the second a high transmutation rate. These options, from Ref 4, are known as the HET2 configuration and the HOM4 configuration.
As a minimum position for the recycling of minor actinides, it would be expected that the ESFR core should be capable of transmuting its own minor actinides. The configuration appropriate to this level of consumption is the HET2 configuration. The HET2 configuration consists of a ring of 84 radial blanket sub-assemblies.
The radial blankets contain fuel composed of 20% MA and 80% depleted uranium. This configuration, with the MA target sub-assemblies outside the active core will have a small impact on the reactivity coefficients of the core.
A high level of minor actinide consumption is provided by the HOM4 configuration. This is capable of transmuting not only the minor actinides produced within the ESFR core but also minor actinides from thermal reactors. The HOM4 configuration consists of 4% by weight of MA homogeneously included within the core fuel, and will have a greater impact on reactivity coefficients than the HET2 configuration.
By choosing the HET2 and HOM4 configurations from the options studied in Ref 4 for the optimised core a wide range of MA transmutation will be considered. These two options covers the full range of impact on the safety performance of including MA transmutation in the reactor and allow judgements to be made as to the safety consequences of MA transmutation.
More detailed analysis of the minor actinide transmutation capabilities of the oxide core for the HET2 and HOM4 configurations, using the CONF2 axial configuration, has been undertaken in the second phase of the work. The impact of MA loading on safety, particularly sodium void reactivity and Doppler constant, in addition to the cores’ efficiency in transmutation of MA has been considered. In addition to the standard configuration of HOM4, which did not include MA in the lower axial blanket, the Ciemat analysis also considered loading of 4% MA into the lower axial blanket of the HOM4 case to further improve MA consumption.
The conclusions to be drawn from this analysis are as follows:
- The HET2 configuration consumes approximately the MA produced by the oxide core. That is the core is approximately neutral in MA production. When this is looked at in more detail there is a small net consumption of Neptunium and Americium and a small net production of Curium.
- The impact on safety of MA in the HET2 configuration, as judged by the sodium void reactivity and Doppler constant, is minimal.
- As expected, the HOM4 configuration has a much higher MA consumption. Over 2050 EFPD, the life of the fuel, the MA transmuted amounts to 870 kg within the active core. Including 4% MA in the lower axial blanket transmutes a further 175 kg. There is, however, a net Curium production in the core and lower blanket of approximately 214 kg.
- Homogeneous loading of the does have a detrimental effect on sodium void reactivity and Doppler constant: sodium void reactivity is increased by approximately 100 pcm and the magnitude of the Doppler constant is reduced by approximately 140 pcm. It should be noted that with the reduction in total sodium void reactivity by the inclusion of the upper plenum of the CONF2 axial configuration and the low enrichment of the fuel in this core, the absolute values of the sodium void reactivity and Doppler constant with homogeneous minor actinide content may still be acceptable. Transient analysis would be required before a final judgment could be made.
Rod Worths
An assessment of the control rod worth of the ESFR oxide core has been undertaken (See Appendix D for details) to ensure that shutdown can be achieved. Assessments of control rod worth for the primary (CSD) and secondary (DSD) CRSs were performed using MCNP and include the determination of the worth curves as a function of rod bank insertion. The assessments were performed for Beginning of Equilibrium Cycle (BOC) condition.
Allowing for a single stuck rod in either the CSD or DSD arrays, the calculated worth of the arrays is as follows:
CSD : 4535 pcm
DSD : 1260 pcm
The requirement to shut down to cold temperatures (100 °C) was calculated and is 2084 pcm.
The CSDs also have to be capable of covering the reactivity loss with burnup, this is approximately 200 pcm, even for a single batch refueling scheme. It is clear that the total requirement of ~2300 pcm can be accommodated by the CSDs with a significant margin when allowing for uncertainty on the calculated values. The DSDs do not need to take up the reactivity loss with burnup since that is done by the CSDs. They do need to cover the reactivity requirement on shutting down to lower temperatures.
Conventionally, the requirement for the CSDs has been to shut down to sodium inlet temperature rather than cold, which would give a requirement of rather less than 2000 pcm. With the array of 9 DSD rods used at present this is not possible. It is judged likely that some transfer of rods from the CSD group to the DSD group would permit the DSDs to meet their requirements without affecting the ability of the CSDs to meet theirs, because of the significant margins available in the current CSD array.
Cores Containing Features for Molten Fuel Discharge
The most efficient options to reduce the void effect, without strongly influencing other parameters, were considered. The studies led to the so called CONF2 configuration. In particular, the axial core structure was modified by the introduction a lower fertile blanket, a larger Na plenum shifted closer to the core, and an absorber layer above the plenum.
A preliminary study of further core design options has been undertaken, related to incorporation of special elements (SAs) with the main aim of facilitating the discharge of molten fuel after a hypothetical severe accident.
The starting point for the study was the CONF2 configuration without MAs. The study provides an evaluation of the neutronics characteristics of these configurations, mainly at BOL conditions, in particular sodium void reactivity and power profiles. The other parameters (Doppler and kinetics parameters, βeff and Λ) are expected to remain practically unchanged.
In particular the following two configurations have been considered:
- CONF3B: where 18 special SAs (wrapper with only Na inside) are introduced in the inner core. 18 fuel SAs have been added to the outer core periphery in order to maintain the total number of fuel SAs.
- CONF3C: the same configuration as CONF-3B, but adopting different special SAs: with the same design as fuel SAs (271 pins, 1.173 cm pitch) but with pins filled with helium instead of fuel.
In both cases, the fuel enrichment for the computation models was the same as in CONF2, as k-eff is not strongly affected by these modifications.
The study has confirmed a high potential of the ESFR core design and, in particular of CONF2, to be a basis for further optimizations. The modifications reduced the core and extended (core and plenum) void effects in CONF3B and CONF3C by about 100 pcm compared with CONF2. If voiding of the special SAs occurs in the case of molten fuel ingress, this also leads to a reactivity reduction. A benchmark against MCNP has shown that the MCNP results are more conservative than the ERANOS ones, except for voiding special SAs. A higher void effect reduction is possible due to incorporation of a larger number of special SAs.
The adoption of special SAs leads to minor power profile variations at BOL. These variations can be reduced by increasing slightly the inner core enrichment. The Power profile at EOEC remains almost the same.
Results of transient analyses show that incorporation of special SAs may improve the core safety performance. Larger effort is needed to get a better understanding on the consequences of the introduction of special SAs under severe accident conditions.
The Carbide Core
The core has a power output of 3600 MWth. Axially the core has a fissile height of 0.8m and contains upper and lower axial blankets.
The fuel sub assembly consists of 331 pins clad in ODS steel with an outer diameter of 8.0 mm and containing a carbide fuel pellet of 6.87 mm outside diameter. The fuel pins are surrounded by a 4.5 mm thick hexagonal EM10 wrapper tube of 178.7 mm outside across flats. The sub-assembly pitching is 178.7 mm.
The performance of the carbide Working Horse core is rather different from that of the oxide core: it has a significantly higher reactivity loss with burnup and a much lower internal breeding gain.
Modifications to produce a more optimised performance of the carbide core are discussed in the following sub-section.
Initial Optimisation of the Carbide Core
The carbide working horse core is a less mature design than the oxide core and required further optimisation. The limits on fuel residence time were defined in the working horse core by the maximum clad damage permitted by ODS, which is 200 dpa. This leads to a peak fuel burnup in excess of 20%, which is rather optimistic for a type of fuel for which there is much less irradiation experience than for oxide. This should be considered to be an interim position until more information becomes available on carbide fuel performance.
The particular criteria for optimisation of the carbide core were as follows:
1. The reactivity loss between start and end of cycle should be reduced to a low level to limit the amount of reactivity that can be injected into the core on accidental withdrawal of one or more control rods.
2. The internal breeding gain of the core should be increased to a level similar to that of the oxide core (~0.02). This is consistent with Criterion 1.
3. Reduce the peak burnup of the fuel to ~15%, as already discussed.
4. Retain the safety advantages of the carbide core which arise from the high thermal conductivity of the fuel, i.e. normal operational fuel temperatures should be lower than those of the oxide core. This can be achieved by ensuring that peak linear ratings do not exceed those of the oxide core.
To reduce the large reactivity loss with burnup, the optimised core height was increased from 0.8m to 1.0m (like that of the oxide core) and the fuel pin outer diameter was increased to 8.5 mm. This has allowed the fuel enrichment to be reduced significantly so that the reactivity loss between start and end of equilibrium cycle when using a 3-batch fuel cycle is only 440 pcm.
The reduction of sodium void reactivity, studied particularly for the oxide core, would be expected to be equally applicable to the carbide core. The CONF-2 axial configuration has therefore been adopted for the carbide core.
The advantages to using carbide fuel arise from its higher density and greater thermal conductivity than oxide. The higher fuel pellet density permits low enrichment to be attained with a smaller fuel pin diameter than the oxide core. The higher thermal conductivity provides greater margin to fuel melting temperature in fault transients, provided that normal operation linear ratings are maintained at a similar level to those of the oxide core. Both of these advantages have been retained in the optimised core design.
The specification of the optimised core is given in Table 1, where the design parameters are compared with those of the carbide Working Horse core. The optimised core plan is given in Figure 3. Because the subassembly pitch has been reduced slightly, minor modifications have been made to the CSD and DSD rods.
The revised dimensions have been determined by scaling of the carbide Working Horse core control rod dimensions.
Rod Worths
An initial assessment of the control rod worth of the ESFR carbide core has been undertaken to ensure that shutdown can be achieved. Assessments of control rod worth for the primary (CSD) and secondary (DSD) CRSs were performed using MCNP and include the determination of the worth curves as a function of rod bank insertion. The assessments were performed for Beginning of Equilibrium Cycle (BOC) condition.
Allowing for a single stuck rod in either the CSD or DSD arrays, the calculated worth of the arrays is as follows:
CSD : 4875 pcm
DSD : 2046 pcm
The requirement to shut down to cold temperatures (100 °C) was calculated and is 1585 pcm.
The CSDs also have to be capable of covering the reactivity loss with burnup. This is approximately 650 pcm, even for a single batch refueling scheme. It is clear that the total requirement of ~2300 pcm can be accommodated by the CSDs with a significant margin even when allowing for uncertainty on the calculated values. The DSDs do not need to take up the reactivity loss with burnup since that is done by the CSDs. They do need to cover the reactivity requirement on shutting down to lower temperatures.
Conventionally, the requirement for the CSDs has been to shut down to sodium inlet temperature rather than cold, which would give a requirement of rather less than 1600 pcm. With the array of 9 DSD rods used at present this is possible, allowing for uncertainties on the calculated values.
Comparison between the Oxide and Carbide Cores
Compared with the Working Horse Carbide Core, the optimized carbide core has undergone significant changes to bring its performance more into line with that of the oxide core. The modifications from the Working Horse Carbide Core have resulted in lower fuel enrichments, nearer to those of the oxide core, with a low reactivity loss over the fuel cycle. The similar performance of the oxide and carbide cores is achieved with a significantly smaller fuel pin diameter and fuel S/A pitch in the carbide core due to the higher density of the carbide fuel. However, at this stage the carbide core has not undergone the same level of optimization as the oxide core.
When comparing safety performance, characterized by sodium void and Doppler constant the carbide core is close to that of the oxide, but not quite as good. CSD absorber rod worths are similar but the DSD array worth is higher in the carbide core than the oxide core.
Whilst fuel temperatures have not been assessed, the carbide core can be judged to have lower fuel temperatures during normal operation, because of the greater thermal conductivity of the carbide fuel, giving larger margins to fuel melt.
Conclusions core optimization activities
Significant progress has been made in optimising both the oxide and carbide ESFR cores.
The oxide working horse core already had a low reactivity loss over the fuel cycle, which has benefits for rod withdrawal faults, so the optimisation process concentrated on the reduction of sodium void reactivity. The CONF-2 axial configuration has provided a significant overall reduction in sodium void reactivity. The working horse oxide core with the CONF-2 axial configuration been adopted as the optimised oxide core.
The carbide working horse core had a significantly higher reactivity loss over the fuel cycle compared with the oxide. By increasing slightly the fuel pin diameter, whilst still retaining the advantages of lower fuel temperatures of carbide fuel, and making some changes to the core layout, the reactivity loss over the cycle has been reduced to a level similar to that of the oxide core. By adopting the CONF-2 axial configuration initially developed for the oxide core, the sodium void reactivity of the carbide core has also been reduced significantly.
Fuel
Nitrides/Carbides fuels
Tests were conducted on the production of (U,Pu,Am)X fuels (X = C and N). the sol gel route was used to produce precursor (U,Pu)O2 beads with carbon in place for the carbothermal reduction. The Pu content Pu/(U+Pu) content was about 20%. Two batches were available with C/(U+Pu) molar ratios in excess of 2.0 and 3.0 for the synthesis of the carbides and nitrides respectively. Both sets of material were infiltrated with an Am nitrate solution, with the beads then being thermally treated under Ar/H2 to convert the Am nitrate to oxide. At this stage their composition can be described as a mixture of (U,Pu)O2, AmO2 and Carbon. Given that the risk of carbon volatisation exists, a slightly higher Am content Am/(U+Pu+Am) than to be found in typical fuels for the homogeneous recycling of minor actinides (i.e. 5%) was chosen. Thus the Am content was of the order of 8%. During the carbothermal synthesis it was expected that this value would diminish slightly, especially for the carbide fuel.
The samples were then removed to a dedicated production facility, where the glovebox atmosphere can reach the required O2 and H2O impurity levels (typically <10 ppm). Carbothermal reduction tests were performed on the (U,Pu) precursors themselves. Good quality carbide and nitride samples have been successfully produced, though a small amount of oxide impurity could always be detected. The corresponding (U,Pu,Am)X samples were also synthesised, and at the time of writing this report are being characterised.
Am-bearing Oxide fuels.
Sol-gel infiltration process and physical property measurements:
The production of the (U,Am)O2 samples was completed earlier in the project using a method based on the infiltration of UO2 beads, with subsequent calsination, compaction and sintering steps. The thermal conductivity of the samples has also been achieved using the JRC- ITU laser flash (LAF) method.
Powder metallurgy process and physical property measurements:
The experimental grid includes four AmBB compositions with two contents of Am (15 and 20%), two O/M ratios (for the 20% Am composition) and two densities (for the 15% Am composition). The two densities correspond to a standard dense microstructure and an optimized one with a low density and a significant part of opened porosity in order to favour gaseous release under irradiation. Three compositions were fabricated and characterized at the ATALANTE facility (CEA Marcoule) during the considered period. It is recalled that the composition with 15% of Americium and the optimized microstructure was fabricated and characterized during the previous period. During this last period, difficulties were encountered when fabricating dense AmBB material. A new method, patented as the UMACS process, was developed successfully for this purpose during the considered period and allowed the fabrication of dense AmBB samples for the CP-ESFR programme. Characterization measurements on fabricated samples include chemical analysis, density measurements and XRD analysis. Samples from the four compositions were sent to the LEFCA laboratory at the CEA Cadarache for physical property measurements. Thermal conductivity and thermal expansion measurements are currently on-going at the LEFCA laboratory. Some samples issued from the composition n°3 (15% Am – dense material) were also sent to the JRC-ITU for melting point and vapour pressure measurements.
Nitride and carbide MA fuel inter-comparison
The in-core behaviour of carbides and nitrides has been assessed based on a literature review of out-of-pile accessible irradiation experiments.
SAFETY
The activities on safety was divided in six areas
Safety objectives and design principles
Technology neutral safety objectives and principles consistent with the European safety framework for new nuclear plants (e.g. EUR, Technical Guidelines considered for EPRTM) were provided. The approach is deterministic and complemented by probabilistic studies. Sodium cooled Fast Reactor (SFR) technology operational and licensing feedback has to be considered and specific risks (in particular, those linked with sodium use) have to be addressed. Basic rules for deterministic safety analyses of internal events, consideration of hazards, safety classification and qualification of components were proposed. The role of Probabilistic Safety Assessments is defined. The design strategy with respect to safety was also adressed: guidelines for the implementation of the main safety functions (reactivity control, decay heat removal, confinement) were provided, then the approach for identification of challenging events (i.e. with a potential of large radiological releases) were applied and relevant possible design strategies proposed with respect to the consideration of whole core accidents.
Defense-in-depth levels and incidents/accidents for Design and Beyond Design Basis events
This activity was focused on the safety goals and general safety principles for future reactors, especially applicable for the ESFR and the characteristics to achieve a robust safety architecture.
For the design of a future reactor, a prerequisite is related to how hypothetical situations of general core meltdown are dealt with. The way to deal these situations has namely to be considered early, because of its prime importance on the design options. So this issue was discussed presenting two possible strategies on this item. The main safety issues related to SFR were analysed on the basis of operating and licensing feedback.
Studies of representative transients and accident scenarios for DB and BDB events
A specific task was devoted to the development multi-physics models for the plant design and the calculation of some postulated plant transients. The task group consisted of eight organisations namely, CEA (France), ENEA (Italy), EDF (France), UPM (Spain), PSI (Switzerland), KIT (Germany), NRG (Netherlands) and JRC-IET (European Commission). Each organisation provided its expertise on the use of system codes for sodium-cooled fast reactor applications. The computational tools, originally developed for commercial light water nuclear reactor safety analyses, have undergone several modifications (coolant material properties, heat transfer and pressure drop model) to be applied for SFR studies. The approach followed in the task was in the first instance to set up a benchmarking exercise where each organisation amended its own system code models of the plant and then compared its calculation results with those of the other partners on a commonly agreed basis. In this respect, this benchmark exercise provided the unique opportunity to compare the performance of the current state-of-the-art system codes for applications related to sodium-cooled fast reactor safety analyses.
The benchmarking activity was followed by the calculation of eight significant transients:
1. Control rod withdrawal
2. Primary pumps acceleration from 30% nominal conditions
3. All secondary pumps trip
4. Total loss of steam generators feedwater
5. Flow bypass of the core
6. Loss of offsite power
7. Partial S/A blockage
8. Pump/core feed duct rupture
Transients for the design extension conditions (DEC) were also investigated
1. Unprotected loss of flow
2. Instantaneous total S/A blockage.
Provisions to mitigate core degradation
A review of sacrificial material candidates has been conducted, considering both neutron absorber and diluent materials. This review was initially based on criteria related to thermophysical and chemical thermodynamic properties (melting and boiling temperature, ability to form a liquid solution with molten fuel…). Neutronic calculations have been done for some generic configurations in order to estimate the reactivity decrease due to the mixing of corium with some different materials. Materials such as aluminium oxide, uranium oxide and hafnium or europium oxides were presented in more details and their relative advantages discussed.
Mass (t) of Eu2O3 needed to avoid re-criticality as a function of non-melted core fraction (%)
Containment and core catcher
The in vessel design of historical and planned SFRs was reviewed with respect to corium relocation paths from the core to the core catcher. The corium source term was assessed based on recent core designs and on preliminary results of SAS-4A calculations, since presently not quantitative information are available on corium masses ejected upwards and downwards from the core. It turned out, that due to the selected core configuration, the ejected corium mass out of the S/A region into the hot plenum is negligible in that sense, but it can be of importance in case of agglomeration of corium. The coolability is strongly dependent on the geometry of the trap and integrity and geometry of the in-vessel structures. Downward relocation of corium was discussed only briefly, depending on the presently available data of the geometry in diagrid and strongback.
The work was based on available data for the core configurations of CP-ESFR, and available results of SIMMER calculations were also considered.
Different mitigation approaches to face a core disruption accident (CDA)
Modelling capabilities
The activity aimed at evaluation of the modeling capabilities of accident scenarios in the frame of the ESFR defence in depth approach using the PIRT (Phenomena Identification Ranking Tables) methodology. The identification of the phenomena that could affect:
- core degradation and in-vessel structure behavior during transient of the design basis domain, potentially leading to local extended core melting accidents;
- releases and transfer of radionuclides under hypothesized accident conditions;
- sodium fire occurrence and evolution;
- sodium chemical reaction occurrence and evolution (with water, air and concrete), was performed as well as evaluation of their knowledge status based on the basis of the past RD work,
The intent was clearly to identify additional research needed for licensing and risk assessment purposes. The importance of each phenomenon was qualitatively evaluated and ranked from a safety point of view as well as the need for additional experimental research. In all, 25 phenomena that are of high importance and have a high need for additional experimental research were identified.
Innovative reactor architectures, components and balance of plant
Plant concept and layouts
Two reactor concepts (one pool and one loop type) where defined and described in detail and then used as starting points for the studies on innovations. Their advantages and drawbacks where compared in the European context. In conclusion, the loop concept would be probably in a lower power range compared to the pool concept due to the high motivation to limit the number of loops and therefore to have large size components. Potential advantages of the loop type (earthquake response, manufacturability in plants, free sodium level constraints) needs to be balanced by the drawbacks linked to gas entrainment and fuel handling issues. Clearly a large investment on loop type issues would be necessary to reach the same level of maturity reached from the much larger European experience on pool type reactors.
Specific studies were devoted to innovative components like the inner vessel, the isolation condensers allowing the improvement the availability and safety by the diversification of decay heat removal systems.
Several alternative designs for the components of the heat transfer system were analysed like kidney shape intermediate heat exchangers allowing a vessel diameter reduction. Integrated pump/exchanger components were considered for cost reduction. Other technologies like electromagnetic pumps and inverted steam generators (sodium circulating inside the tubes) were studied.
Electromagnetic pump with removable outer inductor
Energy conversion systems
Both cycles, water/steam and gas where considered.
The work was focused on the review of incentives and pre-design of Steam Generators minimizing the risks of large sodium water reaction. Considered solutions were double-wall SG with/without leak detection, Inverted type SG with sodium inside tubes, double tube with solid matrix. Concerning inverted SG, a feedback from past experience has been provided as well as design optimization and extrapolation to larger power units.
The evaluation of double component (IHX/SG) technology with coupling fluid was performed for which studies included a comparison between design candidates, a review of the safety and operational implications including corrosion and erosion, risks of plugging and leak detection issues. Methods for SG leak detection as well as ISI including innovative SG were reviewed.
Double component featuring IHX and SG bundles couple by circulating Pb-Bi
Crosscut activities - Tools and experiments in support to the design and codes and standards.
The activity was focused on three areas:
- Review of tools and experiments needs in support of new safety related systems and components
- Review of tools and experiments needs on thermal-hydraulics, mechanics, chemistry
- Update of Codes & Standards in support of ESFR needs including: State of the Art of existing Codes & Standards
R&D needs for primary inner vessel, cover gas and circuits have been evaluated. For the energy conversion system the different steam generator design have been summarised and an R&D map was prepared. Finally, innovative solutions proposed by the ESFR partners for the decay heat removal system, the external vessel storage tank of the fuel handling system and the core catcher have been assessed. The identified R&D roadmap was summarised together with the available facilities and tools.
Links with the European ADRIANA project have been identified: the list of needs to be addressed by R&D, in support to the development of SFR (and more particularly to French ASTRID project), the list of existing facilities, dedicated to SFR development (available or to be refurbished), the list of facilities, to be designed and/or implemented. All these facilities have been ranked, and a road-map has been established. A list of European facilities, criteria for ranking, priorities for investment, first cost evaluations have also been established
Current state of the art in terms of Codes and Standards (RCC-MR, ASME, EN13445, etc) and have identified actions required on the R&D side and on the codification side to respond to specific needs either resulting from new design options or from evolution of regulatory framework (European Pressure Equipment Directive and associated national regulation).
Education and training
Workshops
The following workshops were organized:
- CEA - ESFR Na and other coolants: in-reactor behaviour and safety
- CEA . ESFR SFR: functional analysis and design safety
- KIT Uni-Karlsruhe - Education and Training Workshop on Model Verification & Validation Principles & Application to ESFR
- KIT Uni-Karlsruhe - Education and Training Workshop on Model Calibration & Predictive Estimation Principles & Application to ESFR (these two last workshops were merged in only one)
- UNI Rome - Workshop on ESFR Engineering Aspects
- Universitad Politécnica de Madrid - Workshop on ESFR Reactor Physics and Safety project results.
More than 150 European young engineers and students attended these successful workshops.
First CP ESFR Workshop at CEA Cadarache
Doctoral dissertations
Five doctoral dissertations were carried out:
- Uni-Karlsruhe
Verification and Experimental Validation of ESFR-core thermal-hydraulics codes and models
- NRG
Sodium fire assessment.
- CEA
Sodium-water interaction, in presence of inert gas or not.
- CEA
Experimental and validation of SFR severe accident phenomenology (finished)
- UPM
Development and verification of European (NURESIM-NURISP) pin-by-pin and nodal coupled NK-TH simulation codes and integrated platform with demonstration applications for ESFR core physics and safety analysis
Potential Impact:
As the end-products developed by the project may be used not only for R&D activities but also for industrial applications, many European industry and safety authorities (or technical safety organizations) are contributing to CP ESFR.
Progressively the research activities on sodium cooled fast reactors at European level are strongly coordinated by the project, which will contribute to an optimised use of European resources.
Through education and training programmes, CP ESFR is developing synergies with educational institutions and thus keeping attractive the concerned domain of activity for students and young researchers. This will also contribute to enhance and preserve in a sustainable way the European scientific leadership.
List of Websites:
http://www.cp-esfr.eu/