Long-life Fast Reactor Research Articles

Enhancing the performance of a long-life modified CANDLE fast reactor by using an enriched 208Pb as coolant

The investigation of the utilization of enriched 208Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every 15 years. Uranium Nitride (enriched 15N), 208Pb, and High-Cr martensitic steel HT-9 were employed as fuel, coolant, and cladding materials, respectively. One of the Pb-nat isotopes, 208Pb, has the smallest neutron capture cross-section (0.23 mb) among other liquid metal coolants. Furthermore, the neutron-producing cross-section (n, 2n) of 208Pb is larger than sodium (Na). On the other hand, the inelastic scattering energy threshold of 208Pb is the highest among Na, natPb, and Bi. The small inelastic scattering cross-section of 208Pb can harden the neutron energy spectrum. Therefore, 208Pb is a better neutron multiplier than any other liquid metal coolant. The excess neutrons cause more production than consumption of 239Pu. Hence, it can reduce the initial fuel loading of the reactor. The selective photoreaction process was developing to obtain enriched 208Pb. The neutronic was calculated using SRAC and JENDL 4.0 as a nuclear data library. We obtained that the modified CANDLE reactor with enriched 208Pb as coolant and reflector has the highest k-eff among all reactors. Meanwhile, the natPb cooled reactor has the lowest k-eff. Thus, the utilization of the enriched 208Pb as the coolant can reduce reactor initial fuel loading. Moreover, the enriched 208Pb-cooled reactor has the smallest power peaking factor among all reactors. Therefore, the enriched 208Pb can enhance the performance of a long-life Modified CANDLE fast reactor.

Reducing Irradiation Damage in a Long-Life Fast Reactor with Spectral Softening

Long-life fast reactors receive considerable attention for their potential of using uranium efficiently, and because they can operate for extended periods without refueling. However, the main obstacle to achieving maximum operating times and fuel burnup is the neutron radiation damage that accumulates in the cladding and structural materials. Simulations of metal-fueled high-burnup fast reactors showed that the damage in these reactors’ cladding material reached 200 displacements per atom (dpa) long before the maximum burnup was achieved. One possibility for overcoming this problem is spectral softening, which would reduce the kinetic energy imparted to reactor materials when neutrons collide with them. In this work, we compared the peak irradiation damage in metal- and oxide-fueled fast reactors with that in equivalent reactors containing beryllium in the fuel and reflectors. We showed that the peak damage to the cladding in a metal-fueled reactor was reduced from 273 dpa to 230 dpa when beryllium was included in the core. In an oxide-fueled reactor, the peak damage to the cladding was reduced from 225 dpa to 203 dpa. All four reactors were operated with a core-average burnup of 112 MWd/kg of initial heavy metal (IHM), without reshuffling or refueling, and contained the same initial actinide mass profiles.

Accident Analysis Simulation in Modular 300MWt Gas Cooled Fast Reactor

Safety analysis of 300MWt helium gas cooled long-life fast reactors has been performed. The analysis of unprotected loss of flow(ULOF) and unprotected rod run-out transient overpower (UTOP) are discussed. Some simulations for 300 MWt He gas cooled fast reactors has been performed and the results show that the reactor can anticipate complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations. GCFR relatively has hard spectrum so it has relatively small Doppler coefficient. In the UTOP accident case the analysis has been performed against external reactivity up to 0.002dk/k. In addition the steam generator design has also consider excess power during severe UTOP case..

Feasibility study on ultralong-cycle operation and material performance for compact liquid metal-cooled fast reactors: a review work

This paper reviews the feasibility of ultralong-cycle operation on a compact liquid metal-cooled fast reactor (LMR) firstly by assessing the operation of a long-life fast reactor core and secondly by evaluating material performance in respect to both long-cycle operation and compact-size fast reactor. Many kinds of reactor concepts have been proposed, and LMR and small modular reactor (SMR) are the issued leading technologies for generation four (Gen-IV) reactor system development. The breed-and-burn strategy was proposed as a core burning strategy to operate a long cycle, and it has been evaluated in this paper with two reactor concepts: constant axial shape of neutron flux, nuclide densities, and power shape during life of energy and ultralong cycle fast reactor. In addition, Super-Safe, Small, and Simple and small modular fast reactor, compact LMR concepts, have been simulated to evaluate their long-life operation strategies. For the other practical issues, the materials for fuel, coolant, and structure have been identified and some of them are selected to have their performance optimized specifically for compact LMR with a long-cycle operation. It is believed that this comprehensive review will propose a proper direction for future reactor development and will be followed by the next step research formore » a complete reactor model with the other reactor components.« less

Local blockage analysis of lead-cooled next-generation nuclear power reactors

Local blockage is one of the possible severe accidents which may occur in lead (Pb)- or lead-bismuth (Pb-Bi)-cooled fast reactors due to their high melting temperature. In a local blockage accident, the local temperature increase cannot significantly reduce power level inherently. Therefore, this type of accident should be considered in the core design optimisation of those reactors. In the present study, a local blockage accident analysis of Pb-Bi-cooled long-life fast reactors has been performed. In this analysis, we consider two models. In the first model, we consider that the cross-flow can only come from two nearby channels. In the second model, we consider that the cross-flow can come from four nearby channels. The results show that for the present long-life Pb-Bi reactors which are of medium power density level, local blockage will increase the coolant, cladding and pellet temperatures but still within their limits.

Study on small long-life LBE cooled fast reactor with CANDLE burn-up – Part I: Steady state research

Small long-life reactor is required for some local areas. CANDLE small long-life fast reactor which does not require control rods, mining, enrichment and reprocessing plants can satisfy this demand. In a CANDLE reactor, the shapes of neutron flux, nuclide number densities and power density distributions remain constant and only shift in axial direction. The core with 1.0 m radius, 2.0 m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead–Bismuth is used as coolant. From steady state analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution, etc. The burn-up velocity is less than 1.0 cm/year that enables a long-life design easily. The core averaged discharged fuel burn-up is about 40%.

ODS Steel Clad MOX Fuel-Pin Fabrication and Irradiation Performance in EBR-II

Oxide dispersion strengthened (ODS) steel cladding on mixed-oxide fuel pins offers high-temperature strength and creep rupture resistance properties for a long-life fast reactor. The fabrication and irradiation of the large diameter ODS steel cladding with high-density mixed-oxide pellets were performed at the Argonne National Laboratory Experimental Breeder Reactor II. Two developmental ODS alloys and two fuel-pin designs were used in this study. The tests were continually monitored during irradiation followed by nondestructive and destructive examinations. Both designs were successful and demonstrated the potential of ODS steel as a cladding material.

Accident analysis of lead or lead—bismuth cooled small safe long-life fast reactor using metallic or nitride fuel

Safety analysis of a lead or lead—bismuth cooled small safe long-life fast reactor was performed. It is proposed that the reactor be used in relatively isolated areas, and operated to the end of its life without refueling or fuel shuffling. In the present paper the reactor power and lifetime are set at 150 MWt and 12 years respectively. In order to assume its safety performance, the following accidents without scram were simulated with neutronic-thermal-hydraulic analysis: unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), simultaneous ULOF and UTOP accidents, and simultaneous ULOF, UTOP and unprotected loss of heat sink (ULOHS) accidents. For each type of accident, four types of long-life small reactor (lead cooled metallic fueled, lead cooled nitride fueled, lead-bismuth cooled metallic fueled, and lead—bismuth cooled nitride fueled) were analyzed. It is shown that all the proposed designs can survive these accidents without requiring help from the operator or active devices.