Feasibility study of large MOX fueled FBR core and applicability of various coolants and fuels aimed at the Self-Consistent Nuclear Energy System

The potential of a large MOX fueled fast breeder reactor (FBR) is evaluated with regard to its ability to transmute radioactive nuclides and its safety when incorporated in the self-consistent nuclear energy system (SCNES). The FBR's annual production amounts of selected long -lived fission products (LLFPs), Se-79, Tc-99, Pd-107, I-129, Cs-135 and Sm-151, can be transmuted by using a radial blanket region and part of a lower axial blanket region without any significant impact on the reactor's nuclear and safety characteristics. The other LLFPs are confined in the system. The hazard index level of the LLFPs per one ton of spent fuel from the system after 1000 years is as small as that of a typical uranium ore. To realize self-controllability (passive safety), the proposed FBR core concept employs gas expansion modules and a sodium plenum above the core. To realize self-terminability, even if MOX fuel melting should cause a core compaction, re-criticality of the core can be avoided by a fuel dilution and relocation module. The results show the MOX fueled FBR core has potential applicability to the SCNES. The fundamental applicability of various coolants and fuels is evaluated based on neutron balance toward the final goal of the ideal SCNES. The results show that gas coolant has a potential for increasing the transmutation efficiency of LLFPs. And an improved SCNES with several conventional FBRs and a FP transmutation reactor is also studied.     

Applicability evaluation to a MOX fueled fast breeder reactor for a self-consistent nuclear energy system

The potential for a MOX fueled fast breeder reactor (FBR) is evaluated with regard to its ability to transmute radioactive nuclides and its safety when incorporated in a self-consistent nuclear energy system (SCNES). The FBR's annual production amounts of selected long-lived fission products (LLFPs), Se-79, Tc-99, Pd-107, I-129, Cs-135 and Sm-151, can be transmuted by using a two layer radial blanket region without a significant impact on core nuclear and safety characteristics. The other LLFPs are confined in the system. The hazard index level of the LLFPs per one ton of spent fuel from the system after 10² years is as small as that of a typical uranium ore. Regarding self-controllability in the system's safety, the proposed FBR core concept has an inherent negative reactivity feedback with a gas expansion module, sodium plenum above the core and burnup reactivity compensation module. So sodium boiling and fuel melting will be avoided in anticipated transient without scram events. Regarding self-terminability, even if the MOX fuel melting should cause a core compaction process, re-criticality of the core can be avoided by a fuel dilution and relocation module.     

Elimination of recriticality potential for the Self-Consistent Nuclear Energy System

The ultimate safety goal of the Self-consistent Nuclear Energy System (SCNES) is to eliminate the recriticality-problem based on a simple safety logic. The principle of the elimination of the recriticality-problem is the Controlled Material Relocation (CMR) to establish the neutronic shutdown by removing the molten fuel to the out of core before a large scale pool formation which has potential of energetics driven by a super prompt criticality.

The CMR concept should be reliable without significant impact on the core neutronic performance. As the typical core concepts to enhance this CMR characteristic, several design options are under consideration. They are fuel assemblies with inner duct structure (FAIDUS), fuel assemblies with hollow fuel pins in the axial blanket region (ABLE) for MOX fueled cores, and fuel assemblies without fuel pin bundle structure in the lower axial blanket region (ELAB) for the metallic fueled core. Based on the core design study and accident analyses, these CMR-oriented concepts have been found feasible without significant degradation of the neutronic performance

In order to experimentally confirm the effectiveness of the CMR concept for the MOX fueled core, the EAGLE project has been started in 1998 by Japan Nuclear Cycle Development Institute (JNC) and The Japan Atomic Power Company (JAPC). The EAGLE project is the experimental program utilizing the out-of pile test facility and in-pile facility IGR of the National Nuclear Center of the Republic of Kazakhstan (NNC/RK).     

Feasibility study on nitride fuel core and recycling system toward Self-Consistent Nuclear Energy System

The concept of a nuclear fuel recycle system with a nitride fueled FBR core has been investigated as a part of related studies towards the Self-Consistent Nuclear Energy System (SCNES). Nitride fuel has been given attention because of its relatively high fuel density and high thermal conductivity. To materialize the SCNES concept, it is important to adequately use the excess neutrons produced in the chain reaction. The high fuel density of the nitride fuel brings out more of the excess neutrons and has a higher potential to transmute the long-lived fission products (LLFP's). The high thermal conductivity, in addition, provides margin of fuel melting, and gives negative feedback due to the Doppler reactivity in unprotected loss of flow accidents. In this paper, we discuss good use of nitride fuel in the SCNES.     

An application of metal fuel cycle technology toward Self-Consistent Nuclear Energy System (SCNES) concept

A fast reactor core and fuel cycle concept has been discussed for Self-Consistent Nuclear Energy System (SCNES) concept. This paper discussed loading material candidates for long-lived fission products (LLFPs) and LLFPs burning capability. Some of LLFPs were possible to be loaded in metal of the generated form. The potential for LLFP-confinement in the reactor system is discussed along with metallic fuel cycle concept. The proposed fuel cycle scheme is a successful candidate for SCNES concept.     

A metallic fuel cycle for Self-Consistent Nuclear Energy System (SCNES)

A fast reactor core and fuel cycle concept has been discussed for Self-Consistent Nuclear Energy System (SCNES) concept. This paper discussed loading material candidates for long-lived fission products (LLFPs) and removal of stable nuclides from radioactive nuclides with isotope separation using tunable laser. Some of LLFPs were possible to be loaded in metal of the generated form. The potential for LLFP-confinement in the reactor system is discussed along with a metallic fuel cycle concept. The proposed fuel cycle scheme is a successful candidate for SCNES concept.     

Effect of TRU fuel loading on core performance and plutonium production of FBR

The effect of trans-uranium (TRU) fuel loading on the reactor core performances as well as the actinide and isotopic plutonium compositions in the core and blanket regions has been analyzed based on the large FBR type. Isotopic plutonium composition of TRU fuel is less than that of MOX fuel except for Pu-238 composition which obtains relatively higher composition. A significant increase of plutonium vector composition is shown by Pu-238 for TRU fuel in the core region as well as its increasing value in the blanket region for doping MA case. Excess reactivity can be reduced significantly (5% at beginning of cycle) and an additional breeding gain can be obtained by TRU fuel in comparison with MOX fuel. Doping MA in the blanket regions reduces the criticality for a small reduction value (0.1%) and it gives a reduction value of breeding ratio. Loading MA in the core regions as TRU fuel composition gives relatively bigger effect to increase the void reactivity coefficient mean while it gives less effect for loading MA in the blanket regions. Similar to the void reactivity coefficient profile, loading MA is more effective to the change of Doppler coefficient in the core regions in comparison with loading MA in the blanket regions which gives slightly less negative Doppler coefficient. Obtained Pu-240 vector compositions in the core region are categorized as practically unusable composition for nuclear device based on the Pellaud's criterion. Less than 7% Pu-240 vector compositions in the blanket region are categorized as weapon grade composition for no doping MA case. Obtaining 9% of Pu-238 composition by doping MA 2% in the blanket regions is enough to increase the level of proliferation resistance for denaturing plutonium based on the Kessler's criterion.     

A feasibility study on FP incineration for self-consistent nuclear energy system (SCNES)

A fast reactor core and fuel cycle concept is discussed for the future “Self-Consistent Nuclear Energy System (SCNES)” concept. The present study mainly discussed long-lived fission products (LLFPs) burning capability and recycle scheme in the framework of metal fuel fast reactor cycle, aiming at the goals for fuel breeding capability and confinement for TRU and radio-active FPs within the system. Combining neutron spectrum-shift for target sub-assemblies and isotope separation using tunable laser, LLFP burning capability is enhanced. This result indicates that major LLFPs can be treated in the additional recycle schemes to avoid LLFP accumulation along with energy production. In total, the proposed fuel cycle is a candidate for realizing SCNES concept.     

Verification of a Nuclear Analysis System for Fast Reactors using BFS-62 Critical Experiment

Critical experiments have been analyzed to verify a nuclear analysis system for fast reactors used in Japan Nuclear Cycle Development Institute (JNC). The experiments were performed in a collaboration work between JNC and the Institute of Physics and Power Engineering of Russia to dispose Russian surplus weapons plutonium, focusing on the effect of the introduction of uranium-plutonium mixed-dioxide (MOX) fuel and stainless steel reflector into the current BN-600 core that is comprised of UO₂ fuel and blanket.

The analysis results agreed well with measured values on most of the nuclear characteristics. The accuracies are comparable to those obtained for the conventional MOX fueled fast reactors. It suggests that the JNC analysis system can analyze accurately nuclear characteristics in uranium fueled cores as well.

A significant improvement was achieved on the sodium void reactivity by employing an ultra fine group cell calculation system. A change in adjoint neutron spectrum mostly contributes to the improvement.

A discrepancy of more than 20% was found on the fission rate distribution of ²³⁵U or ²³⁹Pu in stainless steel reflector regions, which cannot be solved by introducing continuous Monte Carlo calculation or different nuclear data sets.     

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.     

Nuclear and thermal–hydraulic characteristics for an LMR core fueled with 20% enriched uranium metallic fuel

As a part of the core design development of KALIMER (150 MWe), the KALIMER core was initially designed with 20% enriched uranium metallic fuel. In this core design, the primary emphasis was given to realize the metallic fueled core design to meet the specific design requirements; 20% and below uranium enrichment and a minimum fuel cycle length of one year. The core was defined by a radially homogeneous core configuration incorporated with several passive design features to give inherent passive means of negative reactivity insertion. The core nuclear performance based on a once-through equilibrium fuel cycle scenario shows that the core has an average breeding ratio of 0.67 and maximum discharge burnup of 47.3 MWD/kg. When comparing with conventional plutonium metallic fueled cores of the same power level, the present uranium metallic fueled core has a lower power density due to its increased physical core size. The negative sodium void reactivity over the core shows a beneficial potential to assure inherent safety characteristics. The transition from the uranium startup to equilibrium cycle is feasible without any design change. Core nuclear performance characteristics in the present core design are attributed to the specific design requirements of enrichment restriction and fuel cycle length.     

Development of Long Life FBR Fuels with Particular Emphasis on Cladding Material Improvement and Fuel Fabrication

The development of FBR fuel systems with high reliability and long in-core residence capability is required to make the fast reactor economically competitive with other electrical energy sources. PNC program of fuels and materials development has been primarily focused on mixed uranium/plutonium oxide (MOX) fuel with cold-worked 316 stainless steel for the past 20 years. Modified 316 stainless steel with excellent swelling resistance and high creep rupture strength was obtained for cladding and duct of the fast prototype reactor MONJU. Advanced austenitic alloys and high strength ferritic alloys are also being investigated for high burnup fuel assemblies of a long life core in large scale FBRs.

In MOX fuel fabrication technology, extensive progress has been achieved during driver fuel fabrication for the experimental reactor JOYO. A new MOX production facility PFPF has been completed with fully automatic and remote handling systems. This facility serves for MONJU core fuel production. The improvement of fuel fabrication technologies promotes cost reduction, safety operation and security from a physical protection standpoint.     

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 UO₂ 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.     

Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN.

It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the ²⁴¹Pu content in the initial fuel, and the decay heat mainly depends on ²³⁸Pu and ²⁴⁴Cm. The contribution of ²⁴⁴Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum. In addition, from the waste disposal point of view, the characteristics of the heat generation FP elements, the platinum group metals, Mo and the long-lived FPs (LLFPs) were also investigated.     

Plutonium recycling in hydride fueled PWR cores

The classic approach to the recycling of Pu in PWR is to use mixed U-oxide Pu-oxide (MOX) fuel. The mono-recycling of plutonium in PWR transmutes less than 30% of the loaded plutonium, providing only a limited reduction in the long-term radiotoxicity and in the inventory of TRU to be stored in the repository. The primary objective of this study is to assess the feasibility of plutonium recycling in PWR in the form of plutonium hydride, PuH₂, mixed with uranium and zirconium hydride, ZrH_1.6, referred to as PUZH, that is loaded uniformly in each fuel rod. The assessment is performed by comparing the performance of the PUZH fueled core to that of the MOX fueled core. Performance characteristics examined are transmutation effectiveness, proliferation resistance of the discharged fuel and fuel cycle economics. The PUZH loaded core is found superior to the MOX fueled core in terms of the transmutation effectiveness and proliferation resistance. For the reference cycle duration and reference fuel rod diameter and pitch, the percentage of the plutonium loaded that is transmuted in one recycle is 53% for PUZH versus 29% for MOX fuel. That is, the net amount of plutonium transmuted in the first recycle is 55% higher in cores using PUZH than in cores using MOX fuel. Relative to the discharged MOX, the discharged PUZH fuel has smaller fissile plutonium fraction - 45% versus 60%, 15% smaller minor actinides (MA) inventory and more than double spontaneous fission neutron source intensity and decay heat per gram of discharged TRU. Relative to the MOX fuel assembly, the radioactivity of the PUZH fuel assembly is 26% smaller and the decay heat and the neutron yield are only 3% larger. The net effect is that the handling of the discharged PUZH fuel assembly will be comparable in difficulty to that of the discharged MOX assembly while the proliferation resistance of the TRU of the discharged PUZH fuel is enhanced.     

Reaction of Titanium Beryllide with Water Vapor

The Doppler reactivity effect of 238U was measured in simulated MOX fuel using the FCA facility for the purpose of obtaining data on the ²³⁸U Doppler reactivity effect in light-water-moderated MOX fuel and evaluating the prediction accuracy of the current analysis code systems and nuclear data library. Experimental data on the Doppler reactivity effect from room temperature up to 800°C were obtained for a uranium fueled core and mockup cores for MOX-fueled LWR using cylindrical natural-uranium samples. With the use of various samples with various neutron spectra, ²³⁸U Doppler reactivity effects at energies generally in the low range below 1 keV were evaluated. Analyses were performed using the current standard analysis code systems for fast and thermal reactors, with the JENDL-3.3 data library. Both analyses yielded calculated/experimental value (C/E) ratios of 0.96 to 1.06 for the MOX cores, a good agreement within the experimental error, and those in the uranium core were similar.     

Safety characteristics of the SCNES core

The core concept of the Self-Consistent Nuclear Energy System (SCNES) and its safety characteristics have been investigated from the view point of the elimination of recriticality. The recriticality potential can be eliminated based on characteristics of self-controllability to prevent the core damage and self-terminability to limit the propagation of core disruption. These two characteristics are simultaneously achieved by the radial heterogeneous two region core with different height. This core consists of leading and driver zones where hybrid metallic fuels with different melting point are installed. The self-controllability can be achieved by decreased coolant density effect due to the above core sodium plenum at the leading zone. The self-terminability is achieved by the Controlled Material Relocation (CMR), which is essentially the preceding downward in-pin fuel relocation selectively generated at the leading zone. U-Pu-1Zr alloy is used to the leading zone fuel due to lower melting point (900°C) than the driver fuel of U-Pu-10Zr(1100°C). Based on the quantitative investigations, it was emphasized that the recriticality potential can be eliminated by the in-pin fuel CMR even for severe unscrammed events such as a total pump stick for the primary coolant system and a total control rods withdrawal.     

Transmutation of long-lived radioactive waste based on double-strata concept

A considerable attention is directed toward the reduction in the long-term potential hazard by partitioning and transmutation (P-T): separating long-lived nuclides from the waste stream and converting them into either shorter-lived or non-radioactive ones. The effects of higher Pu and minor actinide (MA) compositions on the transmutation rates have been studied for a typical mixed oxide (MOX)-fuel fast breeder reactor (FBR) core with 2600 MWt. The calculations showed that the transmutation rate for (Pu, MA) compositions from MOX -LWR becomes one half than that from UO₂-light water reactor (LWR). Furthermore, MA accumulation and transmutation based on Double-Strata Scenario have been investigated for introducing the accelerator driven transmutation system (ADS) with 800 MWt. It was shown that in the scenario of nuclear plant capacities for maximum 140 GWe, which consists of LWRs and FBRs, the introduction of ADS can play a significant role as “Transmuter” in the back-end of fuel cycle.     

Extension of core life-time in fast breeder reactor by introducing inner blanket and doping minor actinides

In order to achieve a longer-life Fast Breeder Reactor (FBR) compared with conventional one, the feasibility study based on proto-type large scale sodium cooled FBR has been performed by utilizing a characteristic of a fertile material of minor actinides (MA) and an inner blanket arranged radially at the center of the core. The analytical results showed that the long-life core without the inner blanket could be achieved by doping MA into an active core because ²³⁸Pu transmuted from MA worked as the fissile material. In case of the core with the inner blanket, it was found that if MA is doped into the inner blanket, the longer-life core also could be achieved by shifting of the main fission reaction zone geometrically from the active core to the inner core due to producing of ²³⁸Pu in the inner blanket. It was also found that if MA is doped into both the inner blanket and the active core, the core life can be extended further. As for the safety characteristics, it has been confirmed that the sodium loss reactivity is improved in case of introducing the inner blanket due to the enhancement of neutron leakage. It has also been confirmed that the sodium loss reactivity is largely affected if the region of high neutron flux, that is the region of main fission reaction is voided.