Radial Blanket Research Articles

Studies on use of reflector material and its position within FBR core for reducing U232 content of U produced in ThO2 radial blankets

Presence of U232 in U233 bred in thorium blanket of fast reactor is a major concern in fuel reprocessing. The former's daughter products being hard gamma emitter and the isotope itself having substantial half life, its presence beyond 10ppm makes fuel recycle complicated and expensive. In this study possibility of decreasing U232 production in a typical FBR blanket by means of spectrum modification is examined. SS, depleted B4C, SiC, Mo and W regions were introduced between core and radial blanket and evolution of isotopes were studied to arrive at an optimal configuration that satisfies requirements of breeding U233 and lowering U232concentration. SS, B4C, SiC, Mo and W are known to be high temperature material with appropriate stability in harsh fast reactor environment. Study has shown that introducing two SS reflector rows can achieve the required low value of U232concentration without greatly compromising the U233production.

Multiple recycling of fuel in prototype fast breeder reactor in a closed fuel cycle with pressurized heavy-water reactor external feed

A fast breeder reactor (FBR) closed fuel cycle involves recycling of the discharged fuel, after reprocessing and refabrication, in order to utilize the unburnt fuel and the bred fissile material. Our previous study in this regard for the prototype fast breeder reactor (PFBR) indicated the possibility of multiple recycling with self-sufficiency. It was found that the change in Pu composition becomes negligible (less than 1%) after a few cycles. The core-1 Pu increases by 3% from the beginning of cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th by only 0.3%. In this work, the possibility of multiple recycling of PFBR fuel with external plutonium feed from pressurized heavy-water reactor (PHWR) is examined. Modified in-core cooling and reprocessing periods are considered. The impact of multiple recycling on PFBR core physics parameters due to the changes in the fuel composition has been brought out. Instead of separate recovery considered for the core and axial blankets in the earlier studies, combined fuel recovery is considered in this study. With these modifications and also with PHWR Pu as external feed, the study on PFBR fuel recycling is repeated. It is observed that the core-1 initial Pu inventory increases by 3.5% from cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th is only 0.35%. A comparison of the studies done with different external plutonium options viz., PHWR and PFBR radial blanket has also been made.

Feasibility study of the fast breeder reactor blanket as a target of plutonium denaturing and TRU transmutation for enhancement of the proliferation resistance and the actinide management in the future fuel cycle

A comparative study was performed with different designs of the fast breeder reactor (FBR) blanket to evaluate the performance of plutonium breeding accompanied with denaturing, and trans-uranium (TRU) transmutation for enhancement of the proliferation resistance and actinide management in the future fuel cycle. Four designs of the FBR blanket fuel were intercompared, conventional depleted uranium (DU) oxide fuels, DU with TRU oxide fuels, DU with TRU oxide fuels with homogeneous/heterogeneous moderators. As results, with heterogeneous moderator in the axial and the radial blanket fuels, the greatest TRU transmutation and plutonium denaturing performance were attained without deteriorating fuel breeding. Design optimization was performed as for ZrH1.65 moderator rod/pellet arrangement in the axial and the radial blanket fuel assemblies. The optimum number of moderator rods per assembly was 25 (11% of the number of total rods) with proper mutual interval (2–3 cm) and symmetrical arrangement in the radial blanket, and insertion of 1 cm thick ZrH1.65 moderator pellets to 4 and 10 cm from the top and the bottom planes of the active core was the optimum design in the axial blanket, in order to achieve the highest TRU transmutation ratio and plutonium denaturing performance without deteriorating fuel breeding. From the non-proliferation aspect, material attractiveness of plutonium were evaluated comparing with the criterion based on decay heat, and required TRU doping ratio in DU with TRU oxide fuel were proved to be as 1.6–2.5 wt.% with heterogeneous moderator, 2.1–3.0 wt.% with homogeneous moderator, and 4.1–5.0 wt.% without moderator. The heterogeneous loading of moderation showed the great advantages in fuel treatment in clear the heat generation limit criterion, as well as mass requirement to the initial and multi-cycle phase.

A neutronic study on advanced sodium cooled fast reactor cores with thorium blankets for effective burning of transuranic nuclides

In this paper, new design concepts of sodium cooled fast reactor (SFR) cores having thorium blanket are suggested for pursuing effective burning of TRU (transuranics) nuclides from LWR spent fuels and their neutronic performances are analyzed. Several core configurations having different arrangements of thorium blankets are explored to improve the core performances and safety-related parameters including sodium void worth which is one of main concerns on safety of SFR cores. Specifically, axial and radial thorium blankets are considered for two type cores. The first one is the typical annular type cores having two different fuel regions where axial thorium blankets are placed in the axially central regions while the second one is the single fuel region cores having central non-fuel region where the axial blanket and radial blankets are considered. Also, the effects of the recycling options and fuel management schemes of the used thorium blanket on the core performances are analyzed. The core performance analyses show that thorium blankets with no recycling option and multi-batch fuel management schemes are very effective to improve the core performances including burnup reactivity swing, sodium void worth and TRU consumption rate.

Development and Application of an Integrated Fuel Performance and Subchannel Model for Analysis of Sodium Fast Reactors

A metal fuel performance code was coupled to a subchannel analysis code to predict, in a computationally efficient way, critical phenomena that could lead to pin failure for steady-state and transient scenarios in sodium-cooled fast reactors. The fuel performance and subchannel codes coupled are FEAST-METAL and an updated version of COBRA-IV-I, respectively. In coupling the codes, the importance of azimuthal temperature and stress effects in the fuel pin were analyzed; it was concluded that azimuthal temperature averaging around the fuel pin is an acceptable approximation. The codes were coupled using a wrapper, the COBRA And FEAST Executer (CAFE), written in the Python programming language. Data from EBR-II was used to confirm and verify CAFE. Finally, CAFE was used to predict the maximum allowable burnup of three different fuel assembly designs (driver fuel, radial blanket, and tight-pitch breed-and-burn fuel) as a function of operating temperature, linear power, fuel composition, cladding thickness, and smear density.

The results of the irradiation experiment MARIOS on americium transmutation

The MARIOS irradiation experiment is the latest in a series of experiments on americium transmutation, and has been carried out in the framework of EURATOM’s 7th Framework Programme (FP7) project FAIRFUELS, which started the 1st January 2009 and is still ongoing. The Post Irradiation Examination (PIE) of MARIOS samples will be performed under the PELGRIMM project. The transmutation of Minor Actinides (MA) is a fundamental step in order to be able to close the nuclear fuel cycle. One of the attractive possibilities to burn MA, is represented by the Minor Actinides Bearing Blanket (MABB) concept. In this option, MA are diluted in a UO2 matrix and irradiated for a long time (from 2000 to 4100days) in radial blankets at the periphery of the outer core of a Sodium Fast Reactor (SFR). Past experimental activities in the field of transmutation and testing of innovative nuclear fuel containing Am has proved that the release or trapping of helium as well as the studies on the swelling of such kind of fuel is a key issue for its safety design. Therefore, the main objective of the MARIOS experiment is the study of the in-pile behaviour of UO2 (with natural uranium) containing Am as minor actinide in order to gain knowledge on the role of the microstructure and of the temperature on the gas release and on fuel swelling for the MABB concept.The MARIOS irradiation experiment started on the 19th of March 2011 and finished on the 2nd of May 2012 after 11 reactor cycles (∼304 full power days) in the HFR (High Flux Reactor) in Petten (The Netherlands). The MARIOS irradiation experiment will determine the performance of the MABB fuel, the behaviour of such targets during the irradiation did not show any anomalies. It is possible to conclude that from an operational point of view, these kinds of fuel which have been developed mainly having in mind the possibility to burn MA in a blanket at the periphery of a Sodium Fast Reactor, did not show significant issues. This paper summarises the main experimental data obtained after the 11-cycle irradiation of the MARIOS experiment in the HFR.

Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus – Spectral indices

The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

Improvement of Core Performance by Introduction of Moderators in a Blanket Region of Fast Reactors

An application of deuteride moderator for fast reactor cores is proposed for power flattening that can mitigate thermal spikes and alleviate the decrease in breeding ratio, which sometimes occurs when hydrogen moderator is applied as a moderator. Zirconium deuteride is employed in a form of pin arrays at the inner most rows of radial blanket fuel assemblies, which works as a reflector in order to flatten the radial power distribution in the outer core region of MONJU. The power flattening can be utilized to increase core average burn-up by increasing operational time. The core characteristics have been evaluated with a continuous-energy model Monte Carlo code MVP and the JENDL-3.3 cross-section library. The result indicates that the discharged fuel burn-up can be increased by about 7% relative to that of no moderator in the blanket region due to the power flattening when the number of deuteride moderator pins is 61. The core characteristics and core safety such as void reactivity, Doppler coefficient, and reactivity insertion that occurred at dissolution of deuteron were evaluated. It was clear that the serious drawback did not appear from the viewpoints of the core characteristics and core safety.

Sodium Experiments of Buoyancy-Driven Penetration Flow into Low-Power Subassemblies in a Sodium-Cooled Fast Reactor During Natural Circulation Decay Heat Removal

Natural circulation plays a significant role in the decay heat removal function of a sodium-cooled reactor. A recent design of the Japan Sodium-Cooled Fast Reactor (JSFR) fully uses natural circulation for a decay heat removal system (DHRS). A dipped heat exchanger (DHX) is immersed in the reactor upper plenum as the DHRS. The DHX provides cold sodium in the upper plenum during the decay heat removal operation. This cold sodium covers the top of the core under the low-flow-rate conditions of natural circulation. Several water experiments of natural circulation in fast reactors revealed that the cold fluid in the reactor upper plenum might partially and temporally penetrate into the low power core channels, e.g., the radial blanket fuel subassemblies. Sodium experiments were carried out to find the onset conditions and the penetration depth of such partial reverse flow driven by buoyancy force. A blanket subassembly and the upper plenum were modeled in the test section including the axial upper neutron shielding of the subassembly. The experimental parameters were the temperature difference between the hot upward flow in the channel and the cold fluid in the upper plenum and the flow velocity in the channel. The onset conditions of the penetration flow were correlated with Gr and Re numbers as well as with basic water experiments. The observed penetration depths were limited to the upper axial neutron shielding of the subassembly.

Approach to equilibrium fuelling scheme of 500 MWe PFBR based on 3-D core burnup modeling

Approach to equilibrium fuelling scheme of 500 MWe prototype fast breeder reactor (PFBR) has been predicted using detailed 3-D core burnup modeling. Equilibrium is reached after two cycles of 180 effective full power days (efpd) each. One-third core is refueled every time in a repeatable scatter load scheme after every 3 cycles. Considering the constraints of linear heat rating (LHR) on fuel and blanket pins it is found that the nominal core achieves full power only in mid-cycle. A novel interpolation scheme is used to find the peak LHR in any axial section of a fuel/blanket sub-assembly. Breeding ratio is adequate for self-sufficient Pu generation in a closed fuel cycle with Pu from axial blankets and two rings of radial blanket sub-assemblies.

Neutronic and nonproliferation characteristics of (PuO 2–UO 2) and (PuO 2–ThO 2) as fast reactor fuels

Computational core physics analysis carried out for a typical fast breeder reactor (FBR) core design is presented through two case studies; one using only (PuO 2–UO 2) MOX fuel and another that replaces ∼41% of (PuO 2–UO 2) MOX with (PuO 2–ThO 2) MOX while conserving the total Pu content. The basic computational framework employed uses the MONTEBURNS2 PERL script to couple the neutronic code, MCNP5 with the depletion/burn-up code, ORIGEN2.2. The parameters computed and compared are: net neutron multiplication factor ( k eff); regionally averaged neutron spectrum; neutron flux; thermal power distribution; breeding ratio; fuel burn-up; fissile material build-up/depletion- 232U build-up; and fuel temperature dependent Doppler Effect, coolant temperature dependent sodium expansion coefficient, and nonproliferation characteristics such as dose rates, spontaneous fission and gamma emissions. The analyses of the case studies indicate that the core physics characteristics, except k eff, are only marginally different in their magnitudes between the two cases, if not equal. The first case study shows that diversion of either 8 radial blanket sub-assemblies (weapon grade Pu) or 1 spent fuel sub-assembly (reactor grade Pu) discharged from an equilibrium core is sufficient to derive a significant quantity (SQ). The second case study shows that a considerable improvement in proliferation resistance can be achieved with the peripheral loading of (PuO 2–ThO 2) MOX pins in all the fuel sub-assemblies of a fast reactor, which should aid in nuclear material safeguards. The comparison of transuranic (TRU) generation for both the cases showed that about 60% reduction in neptunium production has been achieved for the new proposed partially ThO 2–PuO 2 loaded FBR design, whereas other higher TRUs like americium, curium, etc. did not show significant reduction.

Strategy for combining fast growth in electricity generation capacity with the early introduction of the thorium cycle in India

The purpose of this paper is to describe a strategy for combining fast growth in electricity generation capacity, with the early introduction of the thorium cycle in India. Three types of Liquid Metal Cooled Fast Breeder Reactors (LMFBRs) are considered: LMFBR 1, with plutonium/depleted uranium driver zones and depleted uranium blankets; LMFBR 2, with self-sustaining plutonium/depleted uranium driver zones and depleted uranium axial blankets, while the radial thorium blanket produces excess U 233; LMFBR 3, with U233-Th driver zones and Th blankets. The strategy suggested is a combination of 50% LMFBR 1 and 50% of a symbiotic mix of LMFBR 2 and LMFBR 3 all the three using metal fuel. With this strategy, the delay in reaching an installed electrical capacity of 350 GWe is likely to be only about 15 years beyond the approximately 50 years envisaged with only metal fuelled LMFBR 1.

Comparative study of minor actinide transmutation in sodium and lead-cooled fast reactor cores

This paper shows that lead-cooled and sodium-cooled fast reactors (LFRs and SFRs) can preferentially consume minor actinides without burning plutonium, both in homogeneous and in heterogeneous mode. The former approach consists of admixing about 5% of minor actinides (MAs) into uranium–plutonium fuels in the core and using a limited number of thermalising pins consisting of UZrH 1.6. These are needed to keep the negative Doppler feedback larger than the positive coolant reactivity coefficient. Our Monte Carlo burn-up calculations showed that a 600 MW e LFR self-breeder without blankets can burn an average of around 67 kg annually of MAs with a reactivity swing of only about −0.7$ per year. The reactivity swing of a corresponding 600 MW e SFR is more than three times larger due to the poorer breeding and half the critical mass in comparison to the LFR. However, when axial and radial blankets loaded with 10% MAs are added, the SFR burns 25% more MAs (131 kg/yr) and breeds 30% more Pu (150 kg/yr) than an equally sized LFR. When only the blankets are loaded with MAs, the SFR breeds 30% more Pu (198 kg/yr) and still burns about 60 kg a year of MAs. However, in terms of severe accident behaviour, the LFR, with its superior natural coolant circulation and larger heat capacity, has definite advantages.

Core characteristics of fast reactor cycle with simple dry pyrochemical processing

Fast reactor core concept and core nuclear characteristics are studied for the application of the simple dry pyrochemical processing for fast reactor mixed oxide spent fuels, that is, the Compound Process Fuel Cycle, large FR core with half of loaded fuels are recycled by the simple dry pyrochemical processing. Results of the core nuclear analyses show that it is possible to recycle FR spent fuel once and to have 1.01 of breeding ratio without radial blanket region. The comparison is made among three kinds of recycle fuels, LWR UO 2 spent fuel, LWR MOX spent fuel, and FR spent fuel. The recycle fuels reach an equilibrium state after recycles regardless of their starting heavy metal compositions, and the recycled FR fuel has the lowest radio-activity and the same level of heat generation among the recycle fuels. Therefore, the compound process fuel cycle has flexibility to recycle both LWR spent fuel and FR spent fuel. The concept has a possibility of enhancement of nuclear non-proliferation and process simplification of fuel cycle.

Studies on physics parameters of metal (U–Pu–Zr) fuelled FBR cores

For faster growth of nuclear power in India, it is essential to shift to the use of metal-fuels in fast breeder reactors (FBR), which gives a higher breeding ratio (BR) and lower doubling time (DT). Also, future commercialization of the FBR fuel cycle necessitates the use of metallic fuel along with the pyro-process recycling, which can be less costly than oxide fuel reprocessing. Two-dimensional diffusion calculations have been performed to investigate the various physics parameters of metal (U–Pu–Zr) fuelled FBR cores as a function of reactor parameters like reactor power, smear density, zirconium content in the fuel and the number of rows in radial blankets. A 1000 MWe fast reactor with U–Pu fuel (i.e. metal-fuel with no zirconium – which is a theoretical possibility now, due to the lack of irradiation experience) can attain a breeding ratio of 1.61 and a reactor fuel doubling time of 6.6 yrs. Two methods to reduce the sodium void reactivity, which is high and positive in metal-fuelled FBR cores, are suggested.

ICONE15-10450 Evaluation of Core Characteristics of FBR and PWR Loaded with Americium-containing Cermet Fuel

To effectively transmute Americium (Am) in a nuclear reactor, an idea of target fuel with a high content of Am and inert matrix has been proposed by one of the authors. We evaluated the Am transmutation amount and core characteristics, such as power distribution and reactivity effect, in a fast breeder reactor (FBR) and pressurized water reactor (PWR) loaded with the present Am target fuel. The FBR loaded with Am target fuel in radial blanket region could transmute Am amount which a 1000 MWe class light water reactor generates per year. Regarding the PWR, although the same amount of Am could be transmuted, the negative reactivity effect depends on the number of target pins and their locations in a target assembly. But, since its negative reactivity effect can be reduced by optimizing the number of target pins and their locations inside the target assembly, the feasibility of the FBR and PWR loaded with the present target fuel has been proven.

Conceptual Design Study of 180 MWt Small-Sized Reduced-Moderation Water Reactor Core

Conceptual design of a Small-sized Reduced-Moderation Water Reactor (S-RMWR) core, which has the thermal output of 180 MW, the conversion ratio of 1.0 and the void reactivity coefficient of negative value, has been constructed. S-RMWR is a technology demonstration reactor which also conducts material and fuel testing for commercial use of Reduced-Moderation Water Reactor (RMWR) in large-scale power plants. It has a very tight triangular fuel rod lattice and a high coolant void fraction. The RMWR core axially has two short and flat uranium plutonium mixed oxide (MOX) regions with an internal blanket region in between, in order to avoid a positive void reactivity coefficient. The MOX regions are sandwiched between upper and lower blanket regions, in order to increase a conversion ratio. In this small reactor core, leakage of neutrons is expected to be larger than in a large core. Therefore, a core design concept different from that for a large core is necessary. Core burnup calculations and nuclear and thermal-hydraulic coupled calculations were performed in the present study with SRAC and MOSRA codes. MVP code was also used to obtain control rod worth. Because of its large neutron leakage, keeping the void reactivity coefficient negative is easier for S-RMWR than RMWR. Thus, the heights of MOX region can be taller and the plutonium enrichment can be lower than in RMWR. On the other hand, to achieve the conversion ratio of 1.0, radial blanket and stainless steel reflector assemblies are necessary, whereas they are not needed for RMWR.

Parametric neutronic analysis of HCLL blanket for DEMO fusion reactor utilizing vacuum vessel ITER FDR design

Three-dimensional parametric neutronics calculations using the Monte Carlo code MCNP-4C have been performed for a DEMO-type reactor based on the Helium-Cooled Lithium-Lead (HCLL) blanket. The aim of the analysis was to minimize the radial blanket thickness, while ensuring tritium self-sufficiency and to assess the shielding performance of the reactor in terms of the radiation loads to the super-conducting toroidal field (TF) coils. It was found that tritium self-sufficiency can be achieved with a breeder zone thickness reduced to no more than 55 cm at a 6Li enrichment of 90%. Assuming a 6Li enrichment of 60%, a breeder zone thickness of 60 cm is required to achieve the target TBR of 1.10 which is assumed to be sufficient to cover potential tritium losses and uncertainties. With regard to the shielding performance it was found that the design limits for the radiation loads to the TF-coil can be met with radial blanket thicknesses of 75 cm, 60 cm and 55 cm utilizing a two-component shield of Eurofer steel and tungsten carbide between the breeder zone and the vacuum vessel. The blanket variants with larger radial breeder zone show better shielding performances due to the reduced Eurofer shielding material acting as gamma radiation emitter in between the breeder zone and the vacuum vessel. In particular the radiation dose absorbed in the Epoxy insulator was shown to be the most critical quantity in this regard.

Breeding of 242mAm in a Fast Reactor

There is growing interest in the use of 242mAm as a nuclear fuel. Since the thermal absorption cross section of 242mAm is very high (σa = 8950 b), the best way to obtain 242mAm is by the capture of fast or epithermal neutrons in 241Am. As a result, we have considered replacing the radial blanket of a fast reactor, which is usually depleted uranium, with 241AmO2.We chose a 714-MW(thermal) MONJU reactor, and we replaced some of the radial blanket and the outer core assemblies with 10 676 kg of 241AmO2 fuel. We calculated the reactor core by using the MCNP Monte Carlo code.The total amount of 242mAm becomes stabilized after 16 yr, but the enrichment does not. In our calculation, ~7.2% enrichment is obtained after 18 yr. Obtaining higher enrichments might indicate that 242mAm nuclear fuel can be used without further enrichment in many cases.The results presented in this paper are considered an upper limit scenario. In particular the target 241Am loading is not likely to be available soon, but 242mAm production from lesser amounts is easily scaled down proportional to the actual mass irradiated.

Compound Process Fuel Cycle Concept and Core Characteristics

A new fuel cycle concept is proposed in which spent fuels of light water reactor are multi-recycled in a fast reactor core without ordinary reprocessing process but with only simple dry pyrochemical processing. Feasibility of the concept has been studied focusing on the core nuclear characteristics adopting radial heterogeneous core configuration in which both recycled light water spent fuels and fast reactor fuels are loaded. Results of the core survey analyses show that four to six time recycling of light water reactor spent fuels with the burn-up of 300 to 400 GWd/t can be achieved with almost no increase of minor actinide content. Such benefits will be expected in this concept, though further research and development efforts are required, as efficient utilization of nuclear fuel resources, enhancement of nuclear non-proliferation due to the fact that all actinides are recycled all together in a lumped state together with the fact that the concept has no radial blanket fuels which contain high fissile plutonium ratio, besides the possibilities of reduction of environmental impacts due to reduced high level waste, reduction of fuel cycle cost through simplified reprocessing procedure, and suppression of light water reactor spent fuel pile-up.