Development of M5 Cladding Material Correlations in the TRANSURANUS Code: Revision 1

The technical report is based on an earlier research on material properties of the M5 structural material. Complementing this research with new M5 data found in open literature, a set of correlations has been developed for the implementation to the TRANSURANUS code. This includes thermal, mechanical, and chemical (corrosion) properties of M5. As an example, thermal capacity or burst stress correlations have been proposed using the available experimental data. The open literature provides a wide range of experimental data on M5, but for some quantities they are not complete enough to be suitable for the implementation to the TRANSURANUS code. A balanced consideration of similarity of M5 characteristics to those of Zircaloy-4 (Zry-4) or E110 have therefore led to the recommendation to use some of these data selectively also for M5. As such, creep anisotropy coefficients of E110 are recommended to be used also for M5.JRC.G.I.4-Nuclear Reactor Safety and Emergency Preparednes

HIPOS Calculations with MCNPX: First Results

Abstract is not availableJRC.F.3-High Flux and Future Reactor

Study on Unprotected Transients for LFRs of 600 MWe Power and Two Different Core Radii and Heights

This paper investigates the influence of core radii and height on core outlet temperatures during unprotected transients for a 600 MWe (1426 MWth) lead-cooled reactor fuelled by (U,TRU)O2. Two core geometries were examined, one compact core of 2 m height and 4 m diameter, and a pancake core of 1.2 m height and 5 m diameter. UZrH1.6 pins were added into core sub-assemblies in order to improve reactivity coefficients (increase the negative Doppler reactivity feedback and decrease the coolant temperature reactivity coefficient). The pin diameter was 12.5 mm and the pitch-to-diameter ratio was 1.6. Each fuel pin had a 3 mm diameter concentric hole in order to reduce the maximum fuel temperatures. For the thermal hydraulics calculations the computational fluid dynamics code STAR-CD was used, and the neutronic and burn-up performance was evaluated by the Monte Carlo codes. It was found that during an unprotected Loss-of-Flow accident, for which it was assumed that the core power remained constant during the whole transient, the pancake core had a temperature increase of only 100 K, compared to almost 200 K for the compact core. During an unprotected Loss-of-Heat-Sink accident at full power the heat up of the coolant is independent of the core type, i.e. the reactor vessel reaches the fast creep temperature (1173 K) within 400 seconds. A Total-Loss-of-Power accident would lead to a temperature peak of 1080 K two days after accident initiation.JRC.F.4-Nuclear design safet

Improving the Safety of Minor Actinide Burning in Fast Reactors

In most of the European Union's 27 member states, nuclear power is considered to be a risk rather than an advantage. The positive aspects of nuclear power, such as greater energy security and lower greenhouse gas emissions, are cancelled out by fears of terrorist attacks against nuclear facilities, the potential misuse of nuclear material and concerns about the safety of nuclear waste disposal and nuclear installations. To achieve effective synergy between breeding and minor actinide (MA) waste burning in a fuel cycle, fast reactors should be designed to recycle MAs from several (2-3) LWRs/CANDUs. For homogeneous recycling this in turn means that the MA fraction in the fuel should be about 5%, which, in comparison to MA-free cores, nearly halves the negative Doppler reactivity feedback, however, and increases the positive coolant temperature reactivity coefficient. In the present study, we compare the effectiveness of different moderating materials to improve Doppler fuel temperature reactivity feedback and the coolant temperature reactivity coefficient in lead-cooled and sodium-cooled fast reactors (LFRs and SFRs) homogeneously recycling MAs. The materials investigated were hydrides (CaH_2), metallic beryllium, and enriched boron carbide (11^B_4C). The calculations showed that, in homogeneous mode, both LFR and SFR systems could burn about 80 kg of MAs per year, with the burn-up reactivity swing remaining below 2.5$/yr. Hydride moderating pins appeared to be the most effective in improving reactivity coefficients and a Doppler coefficient twice as large as coolant temperature reactivity feedback was obtained in both an LFR and SFR when ~5% of the fuel pins in each sub-assembly were replaced with CaH_2 moderating pins. The hydrides also proved to have the least influence on core breeding performance, but susceptibility to decomposition at relatively low temperatures (~1100 K) might pose safety problems. The moderating power of beryllium and 11^B_4C is significantly smaller than that of hydrides, which means that the number of moderating pins in a sub-assembly has to be higher. This, however, has a negative influence on breeding and further exacerbates the burn-up reactivity swing. The LFR self-breeder shows a number of advantages over an SFR counterpart regarding its behaviour in unprotected loss-of-flow and unprotected loss-of-heat sink accidents due to superior natural coolant circulation and the larger heat capacity of lead.JRC.F.4-Nuclear design safet

Comparison of Lead and Sodium-Cooled Reactors. Safety, Fuel Cycle Performance and Some Economical Aspects

This paper compares the Lead-Cooled Fast Reactor (LFR) and the Sodium-Cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of-Flow, Loss-Of-Heat-Sink, and Total-Loss-Of-Power. It also studies possible moderators, like BeO and hydrides.JRC.F.4-Nuclear design safet

Achieving Sustainability in Fuel Cycles with Th-Fuelled Thermal Breeders

This paper outlines the path to reach sustainable fuel cycles with Th-fuelled thermal breeder reactors. To achieve a successful synergy between radiotoxic waste transmutation from water cooled reactors and breeding of the new fuel in the Th-233U cycle, fast reactors (FRs) are applied to transmute minor actinides (MAs). It is shown that near-term 233U breeding in PWR and FR cores is feasible. Over 100 kg of 233U could be produced annually in PWR cores with 30% of sub-assemblies containing ThO2, which means that one 850 MWe pressurized heavy water moderated thermal self-breeder could be started every 16 years. A slightly higher figure, over 120 kg/yr of 233U, was obtained when 30% of UO2 pins in each sub-assembly were exchanged for ThO2 pins. Fast reactors employing (Th,TRU)O2 fuel produced up to 370 kg of 233U annually, which means that a new self-breeder could be started roughly every 4.4 years. At the same time, 78 kg of minor actinides are consumed annually. By the end of this century, PWRs and FRs could generate enough 233U to sustain an increase in nuclear power capacities to 1160 GWe, which represents more than a three-fold increase of installed nuclear capacities worldwide. By 2200, the amount of TRUs in the fuel cycle could also be decreased and stabilized.JRC.F.4-Safety of future nuclear reactor

Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues

A large number of new fast reactors may be needed earlier than foreseen in the Generation IV plans. According to the median forecast of the Special Report on Emission Scenarios commissioned by the Intergovernmental Panel on Climate Control nuclear power will increase by a factor of four by 2050. The drivers for this expected boost are the increasing energy demand in developing countries, energy security, but also climate concerns. However, staying with a once-through cycle will lead to both a substantially increased amount of high-level nuclear waste and an upward pressure on the price of uranium and even concerns about its availability in the coming decades. Therefore, it appears wise to accelerate the development of fast reactors and efficient re-processing technologies. In this paper, two fast reactor systems are discussed—the sodium-cooled fast reactor, which has already been built and can be further improved, and the lead-cooled fast reactor that could be developed relatively soon. An accelerated development of the latter is possible due to the sizeable experience on lead/bismuth eutectic coolant in Russian Alpha-class submarine reactors and the research efforts on accelerator-driven systems in the EU and other countries. First, comparative calculations on critical masses, fissile enrichments and burn-up swings of mid-sized SFRs and LFRs (600MWe) are presented. Monte Carlo transport and burn-up codes were used in the analyses. Moreover, Doppler and coolant temperature and axial fuel expansion reactivity coefficients were also evaluated with MCNP and subsequently used in the European Accident Code-2 to calculate reactivity transients and unprotected Loss-of-Flow (ULOF) and Loss-of-Heat Sink (ULOHS) accidents. Further, ULOFs as well as decay heat removal (protected Total Loss-of-Power, TLOP) were calculated with the STAR-CD CFD code for both systems. We show that LFRs and SFRs can be used both as burners and as self-breeders, homogeneously incinerating minor actinides. The tight pin lattice SFRs (P/D = 1.2) appears to have a better neutron economy than wide channel LFRs (P/D = 1.6), resulting in larger BOL actinide inventories and lower burn-up swings for LFRs. The reactivity burn-up swing of an LFR self-breeder employing BeO moderator pins could be limited to 1.3$ in 1 year. For a 600MWe LFR burner, LWR-to-burner support ratio was about two for (U, TRU)O2-fuelled system, while it increased to approximately 2.8 when (Th, TRU)O2 fuel was employed. The corresponding figures for an SFR were somewhat lower. The calculations revealed that LFRs have an advantage over SFRs in coping with the investigated severe accident initiators (ULOF, ULOHS, TLOP). The reason is better natural circulation behavior of LFR systems and the much higher boiling temperature of lead. A ULOF accident in an LFR only leads to a 220K coolant outlet temperature increase whereas for an SFR the coolant may boil. Regarding the economics, the LFR seems to have an advantage since it does not require an intermediate coolant circuit. However, it was also proposed to avoid an intermediate coolant circuit in an SFR by using a supercritical CO2 Brayton cycle. But in an LFR, the reduced concern about air and water ingress may decrease its cost further.JRC.F.4-Nuclear design safet

Severe Accident Considerations for the Lead-Cooled Fast Reactor

The Lead-Cooled Fast Reactors (LFR) has in some cases advantages from a safety point of view compared to other fast reactors. Conversely, the LFRs also have some unresolved issues that need further investigation. Among the advantages is the high coolant boiling temperature of 1780°C, which makes voiding of the core from coolant boiling very unlikely. Since the LFR has low operating pressures, the risk for Loss-Of-Coolant Accidents is reduced. The natural circulation capacity of lead is excelent, which is useful during loss of forced pumping power. The natural circulation is futher improved by that the LFRs usually have a low-pressure drop core with relatively large pitch-to-diameters (PTD) and simple flow paths. For example, some smaller LFR designs, as the STAR-LM [1], by employing this achieves a purely natural circulation cooled core during normal operation. Unresolved issues for LFRs match often the category of material issues. Today the highest temperature that steel, with aluminun coating of the GESA method, can endure for longer times is 600°C and lead velocities should not exceed 2.5 m/s. The most critical structures are the pumps (high velocities), the core and the heat-exchangers (HX). Alternative steels are being considered like for example the MATHAL (Ti3SiC2), but their performance needs to be proven in the lead environment at high temperatures and velocities for longer times. Another uncertainty concerning the LFR is the consequence of a steam generator tube rupture. High-pressure steam would then enter the primary circuit. The strenth of the pressure wave is uncertain as well as where the steam bubbles would move. In case the bubbles reach the core region positive reactivity would increase its power. Research began on LBE and lead cooled Accelerator-Driven Systems (ADS) about ten years ago within European projects called PDS-XADS and later IP EUROTRANS [3]. The latter concerns the developmend of a European Facility Industrial Transmutation (EFIT). This paper compares the performance of the EFIT reactor during the Unprotected Loss-of-Flow (ULOF) accident.JRC.F.4-Nuclear design safet

Lead-Cooled Fast Reactors with Th-based Fuels - Neutronics and Safety

In this paper, neutronic and safety performance of 600 MWe lead-cooled fast reactors (LFRs) employing Th-based fuels is studied. This includes investigation of both (self-) breeders fuelled by Th-233U and burner/breeder reactors incinerating Pu and transuranic elements (TRUs) from spent LWR fuel. Both systems use mixed oxide fuel. Neutronic and depletion characteristics were evaluated by the Monte Carlo code MCB. The European Accident Code-2 (EAC-2) and the Computational Fluid Dynamics code STAR-CD were used for safety analyses. The accidents studied were unprotected Loss-of-Flow, unprotected Loss-of-Heat Sink, and protected Total Loss-of-Power. Our calculations indicate that LFRs can be run in pure fast Th-233U mode. This is, however, at the expense of having large core actinide masses exceeding 100 tons at BOL.(Th,TRU)O2-fuelled LFRs can annually incinerate up to 320 kg of plutonium and about 80 kg of minor actinides. The latter corresponds to an annual MA production in about 1.6 EPR reactors. At the same time, approximately 225 kg of 233U are produced, which can be used to start-up new fast reactors in a pure Th-233U mode or fuel LWRs or advanced thermal PHWR or MSR breeders. A use of axial and radial Th blankets with 5% of MAs further increases the amount of 233U bred in (Th,TRU)O2 system to 285 kg/y. We confirm that requirements for 233U enrichments in fast systems employing Th-based fuels are lower than for LWR-Pu and LWR-TRU fuelled cores. Regarding the safety performance, in-core moderating pins (CaH2) are used to improve Doppler and coolant temperature reactivity coefficients. Finally, we also discuss the favorable behavior of our LFR core designs in postulated severe accidents.JRC.F.4-Nuclear design safet