The closed on-site fuel cycle of the brest reactors

The BREST fast reactor with nitride fuel and lead coolant is being developed as a reactor of new generation, which has to meet a set of requirements placed upon innovative reactors, namely efficient use of fuel resources, nuclear, radiation and environmental safety, proliferation resistance, radwaste treatment and economic efficiency. Mixed uranium-plutonium mononitride fuel composition allows supporting in BREST reactor CBR≈1. It is not required to separate plutonium to produce “fresh” fuel. Coarse recovered fuel purification of fission products is allowed (residual content of FPs may be in the range of 10⁻² – 10⁻³ of their content in the irradiated fuel). High activity of the regenerated fuel caused by minor actinides is a radiation barrier against fuel thefts. The fuel cycle of the BREST-type reactors “burns” uranium-238, which must be added to the fuel during reprocessing. Plutonium is not extracted during reprocessing being a part of fuel composition, thus exhibiting an important nonproliferation feature.

The radiation equivalence between natural uranium consumed by the BREST NPP closed system and long-lived high-level radwaste is provided by actinides (U, Pu, Am) transmutation in the fuel and long-lived products (I, Tc) transmutation in the blanket. The high-level waste must be stored for approximately 200 years to reduce its activity by the factor of about 1000.

The design of the building and the entire set of the fuel cycle equipment has been completed for the demonstration BREST-OD-300 reactor, which includes all main features of the BREST-type reactor on-site closed fuel cycle.     

Non-proliferation criteria for nuclear fuel cycle options

We compared the implied meaning of proliferation resistance in proposals regarding the nuclear fuel cycle. We find discrepancies amongst the proposals regarding the technical definition of proliferation resistance, however there is a consistent focus on the importance of “physical form” as a key in determining a proliferation resistant fuel cycle. The reviewed proposals make little explicit mention the importance of the time to process the material to construct a nuclear weapon. While certain proposals discuss the importance of non-separation of plutonium from actinides, there are other proposals, which clearly do not view this aspect as vital in defining proliferation resistance. Recognizing that there are numerous political and infrastructure measures that may also be taken to guard against proliferation risks, we have focused here on the definition of proliferation resistance in terms of technical characteristics.     

Improvement of plutonium proliferation resistance by doping minor actinides

This paper focuses on improving the proliferation resistance of plutonium resulting from uranium-based fuel irradiation. Intrinsic properties of plutonium isotopes with even mass numbers (²³⁸Pu, ²⁴⁰Pu and ²⁴²Pu) — in terms of their intense decay heat and high spontaneous fission neutron rates — were used as a measure to improve the proliferation resistance of plutonium itself. The present study explores MA addition effect into LEU (5%²³⁵U) and HEU (20%²³⁵U) with regard to plutonium proliferation resistance characteristics. Consideration goes beyond critical condition to examine the potential of subcritical system in enhancing the plutonium proliferation properties. Results show that even the doping level of 1% of Np, TrPu or all MA elements into low enriched uranium improves the proliferation-resistant properties of plutonium. A potential for further improvement is achieved by higher doping of minor actinides into high enriched uranium irradiated in a subcritical mode.     

Proposal of a plutonium burner system based on HTGR with high proliferation resistance

An innovative plutonium burner concept based on high temperature gas cooled reactor (HTGR) technology, “Clean Burn”, is proposed by Japan Atomic Energy Agency (JAEA). That is expected to be as an effective and safe method to consume surplus plutonium accumulated in Japan. A similar concept proposed by General Atomics (GA), Deep Burn, cannot be introduced to Japan because of its adopting highly enriched plutonium, which shall infringe on a Japanese nuclear nonproliferation policy according to Japan–US reprocessing negotiation. The Clean Burn concept can avoid this problem by employing an inert matrix fuel (IMF) and a tightly coupled fuel reprocessing and fabrication plants. Both features make it impossible to extract plutonium alone out of the fabrication process and its outcomes. As a result, the Clean Burn can use surplus plutonium as a fuel without mixing it with uranium matrix. Thus, surplus plutonium alone will be incinerated effectively, while generation of plutonium from the uranium matrix is avoided. High neutronic performance, i.e., achievement of burn-up of about 500 GWd/t and consumption ratio of plutonium-239 reaching to about 95%, is also assessed. Furthermore, reactivity defect caused by the inert matrix is found to be negligible. It is concluded that the Clean Burn concept is a useful option to incinerate plutonium with high proliferation resistance.     

Uranium-based TRU multi-recycling with thermal neutron HTGR to reduce environmental burden and threat of nuclear proliferation

To reduce environmental burden and threat of nuclear proliferation, multi-recycling fuel cycle with high temperature gas-cooled reactor has been investigated. Those problems are solved by incinerating trans-uranium (TRU) nuclides, which is composed of plutonium and minor actinoid, and there is concept to realize TRU incineration by multi-recycling with fast breeder reactor. In this study, multi-recycling is realized even with a thermal reactor by feeding fissile uranium from outside of the fuel cycle instead of breeding fissile nuclide. In this fuel cycle, recovered uranium and natural uranium are enriched and mixed with recovered TRU to fabricate fresh fuels.

The fuel cycle was designed for a gas turbine high temperature reactor (GTHTR300). Reprocessing is assumed as existing reprocessing with four-group partitioning technology.

As a result, the TRU nuclides excluding neptunium can be recycled by the proposed cycle. The duration of potential toxicity decaying to natural uranium level can be reduced to approximately 300 years, and the footprint of repository for high-level waste can be reduced by 99.7% compared with the standard case. Surplus plutonium is not generated by this cycle. Moreover, incineration of TRU from light water reactor cycle can be performed in this cycle.     

Effect of decontamination factor of recycled actinide and FP on the characteristics of SCNES

Effect of the decontamination factor (DF) of nuclides to be confined in the closed fuel cycle was examined in terms of the system characteristics. Two kinds of indices, equilibrium mass and equivalent radiotoxicity, were used to determine the recovery perfectivity of the nuclides. By using the equilibrium mass, extremely high DF had to be attained. The required values of DF were 10⁸ for LLFP and 10¹² for actinides. On the other hand, using the radiotoxicity, inferior perfectivity of recovery, DF=10⁶, could be acceptable for actinides and there is no necessity of LLFP recycling for transmutation because they only have comparable radiotoxicity with the supplied uranium to the system. The required DF which provide ignorable loss of waste to the outside of the system depends on what index we use. Generalization of index to quantify the hazard or nuisance of nuclear waste and setting the criteria remains unsettled questions.     

Actinide reformer concept

The partitioning and transmutation technology is effective to reduce the environmental impact from disposition of high-level radioactive wastes and improve the efficiency of geological disposal. However, Am and Cm and their daughter nuclides are difficult to handle in the fuel cycle because of their high decay heat and radioactivity. These nuclides also give the chemical instability which harms the soundness of fuel pellet properties.

We propose a new system concept “actinide reformer”, which reforms Am and Cm into Pu by neutron capture reactions and decay. Am and Cm are separated from the PUREX reprocessing process and converted to chloride molten-salt fuel. Using liquid-type fuel, above mentioned defects can be compensated. Actinide reformer is an accelerator-driven system which is composed of a 10 MW-class cyclotron, a tungsten target and a subcritical core. Spent molten-salt fuel is reprocessed as an on-line fuel exchange manner to extract fission products and recover Pu to send back to a power generation cycle. The decay heat and neutron emission from the fuel with recovered Pu are smaller than those of MOX fuel with 5% of minor actinide addition. It expects no significant engineering difficulties and cost increase for construction of MOX fuel based reprocessing/fabrication plant and power reactors.     

Optimum utilization of nuclear fuel with gas and vapor core reactors

Gas and Vapor Core Reactors (G/VCR) are externally reflected and moderated nuclear energy systems fueled by stable uranium compound in gaseous or vapor phase. In G/VCR systems the functions of fuel and coolant are combined and the reactor outlet temperature is not constrained by solid fuel-cladding temperature limitations. G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Furthermore, G/VCR systems feature a low inventory and fully integrated fuel cycle with exceptional sustainability and safety characteristics. With respect to fuel utilization, there is practically no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from almost solely a burner to a breeder is achievable. The continuous recycling of fuel in G/VCR systems allows for continuous burning and transmutation of actinides without removing and reprocessing of the fuel. The only waste product at the backend of the gas core reactors' fuel cycle is fission fragments that are continuously separated from the fuel. As a result the G/VCR systems do not require spent fuel storage or reprocessing.

G/VCR systems also feature outstanding proliferation resistance characteristics and minimum vulnerability to external threats. Even for comparable spectral characteristic, gas core reactors produce fissile plutonium two orders of magnitude less than Light Water Reactors (LWRs). In addition, the continuous transmutation and burning of actinides further reduces the quality of the fissile plutonium inventory. The low fuel inventory (about two orders of magnitude lower than LWRs for the same power generation level) combined with continuous burning of actinides, significantly reduces the need for emergency planning and the vulnerability to external threats. Low fuel inventory, low fuel heat content, and online separation of fission fragments are among the key constituent safety features of G/VCR systems.     

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.     

硝酸羟胺用于钚纯化循环的工艺条件优化

采用可控温的单级萃取装置，对羟胺还原反萃取钚的工艺条件进行了优化。实验表明，硝酸肼能够将少量Pu(Ⅳ)还原反萃取到水相，但是当硝酸肼浓度较高时，硝酸肼则表现出盐析效应，抑制钚的还原反萃取；对于钚还原反萃取工艺来说，当保持进料中羟胺与钚的摩尔数之比为定值时（在50℃时n (HAN)/ n (Pu)=2～3较为适宜），增大还原剂流量能够提高钚的收率，但同时会降低钚的浓缩倍数；温度升高时，硝酸氧化Pu(Ⅲ)的反应速率加快，使得钚在有机相中的浓度有所升高；当溶液中离子强度较高时，在盐析效应的作用下，Pu(Ⅲ)的分配比随离子强度的提高而升高，导致钚在有机相中的浓度上升。     

Double-strata high burnup fuel performance in light water reactors

This study is aimed at the development of a fuel cycle concept for host countries with a lack of nuclear infrastructure. Two interrelated criteria, proliferation resistance and high-burnup, form the general framework of the fuel management scenario with the highest priority given to light water reactor technology and plutonium-free fresh fuel. Logically it implies the use of uranium oxide with enrichment close to 20%, whose effective utilization forms the main subject of the present paper. A sequence of two irradiation cycles for the same fuel pins in two different light water reactors is the key feature of the advocated approach. It is found that the synergism of PWR and pressure tube graphite reactor offers fuel burnup up to 140 GWd/tHM. Being as large as 8% in the final isotopic vector, the fraction of ²³⁸Pu serves as an inherent protective measure against plutonium proliferation.     

Evaluation of implementation of thorium fuel cycle with LWR and MSR

In order to construct a sustainable society, it is necessary to consider fairness beyond generations and between countries. It is expected that Asian countries continue growing their economy and will result consuming more energy. More CO₂ emission is not acceptable.Nuclear power has many advantages for reducing CO₂ emission. However, it still has concerns of nuclear proliferation, radioactive waste and safety. It is necessary to overcome these concerns if nuclear power is expanded to Asian countries. Thorium utilization as nuclear fuel will be an opening key of these difficulties because thorium produces less plutonium, less radioactive waste. Safety will also be enhanced. The use of molten-salt reactor (MSR) triggered by plutonium supply from ordinary light water reactor (LWR) with uranium fuel will allow implementation of thorium fuel cycle with electricity capacity of about 446 GWe around at 2050.The other important sector in a view of sustainability is transportation. Transportation is essential for economy growth. Therefore it is inevitable to reduce CO₂ emission from transportation sector. Electric vehicle (EV) will be used as a major mobility instead of gasoline engine cars. Rare-earth materials such as neodymium and dysprosium are necessary for producing EV. These materials are expected to be mined from Asian countries. It is often obtained with thorium as by-product. Thorium has not been used as nuclear fuel because it is not good for nuclear weapon and it does not have fissionable isotopes. Recent global trend of nuclear disarmament and accumulation of plutonium from uranium fuel cycle can support starting the use of thorium.Thorium utilization will help both to provide clean energy and to produce rare-earth for clean vehicle. These will create new industries in developing Asian countries. An international collaborative framework can be established by supplying resource from developing countries and supplying technology from developed countries. “THE Bank (THorium Energy Bank)” is proposed here as one part of such a framework.     

Specific fuel rod thermomechanical studies for the Advanced Plutonium fuel assembly concept

In order to burn plutonium in PWRs, an innovative assembly has been proposed: the Advanced Plutonium fuel Assembly (APA) concept. This heterogeneous assembly is based on the standard 17x17 geometry, replacing 144 classical UO₂ fuel rods by 36 large fuel rods containing plutonium in an inert matrix. To accommodate the high power and burn-up, a thin annular geometry has been proposed. However, the design of such assembly is a challenge due to the highly innovative fuel rod concept. Regarding the fabrication, different processes are under investigation: concentric claddings with annular ceramic pellets (cercer), ceramic/metal fuel (cermet) metallurgically bonded to the claddings, vibrocompacted fuel. This paper focuses on thermomechanical studies performed for the two first concepts, in view of a future selection between the different processes. The analysis concerns temperature profiles, fuel-cladding gap evolutions and cladding buckling risk. Regarding those parameters, the conclusion is that the cermet matrix presents satisfactory conditions: operating at very low temperatures and no buckling risk.     

The encapsulated nuclear heat source reactor for low-waste proliferation-resistant nuclear energy

The encapsulated nuclear heat source (ENHS) is a new Pb-Bi cooled modular reactor concept that features a combination of the following useful features that may make nuclear energy more attractive: (1) 20 years of full power operation without refueling. (2) Nearly constant fissile fuel contents and k_eff. (3) No on-site refueling and fueling hardware. (4) The ENHS modules are factory manufactured and transported already fueled to the site. (5) No access to neutrons. (6) No mechanical connections between the ENHS module and the energy conversion plant (The ENHS module has the function of a nuclear battery — with 20 years of full power operation at 125 MW_th). (7) At end of life, the ENHS module serves as a spent fuel storage cask and, later, as a spent fuel shipping cask. That is, the fuel is locked inside the ENHS from “cradle to grave”. (8) 100% natural circulation resulting in passive load following capability and autonomous control. This combination of features offers a highly safe nuclear energy system that is characterized by low waste, high proliferation resistance and high uranium utilization. The low waste and high uranium ore utilization are achieved by recycling the Pu and MA many times using a proliferation-resistant dry process; only fission products are to be extracted between cycles. Spent LWR fuel can provide for the HM make-up. The high level of proliferation resistance is obtained by restricting access to the fuel and neutrons and by eliminating the economic incentive of the client country to invest in sensitive technologies or infrastructure that can be used for clandestine production of strategic nuclear materials.     

Actinide recycling by pyro-process with metal fuel FBR for future nuclear fuel cycle system

Pyro-metallurgical technology is one of potential devices for future nuclear fuel cycle. Not only economic advantage but also environmental safety and strong resistance for proliferation are required for the fuel cycle. In order to satisfy the requirement, actinides recycling applicable to LWR and FBR cycles by pyro-process has been developed since more than ten years in CRIEPI. The main technology is electrorefining for U and Pu separation and reductive-extraction for TRU separation, which can be applied on oxide fuels through reduction process as well as metal fuels. The application of this technology on separation of TRU in HLLW through chlorination could contribute to the improvement of public acceptance on the geologic disposal.

The main achievements are summarized as follows:

• -|The elemental technologies, such as electrorefining, reductive extraction, injection casting and salt waste treatment and solidification, have been developed successfully with lots of experiments

• -|The fuel dissolution into molten salt and uranium recovery on solid cathode for electrorefining have been demonstrated by engineering scale facility in Argonne National Laboratory by using spent fuels and in CRIEPI by uranium tests.

• -|Single element tests, using actinides, showed the Li reduction to be technically feasible, remaining the subjects of technical feasibility on multi-elements system and on effective recycle of Li by electrolysis of Li₂O.

• -|Concerning on the treatment of HLLW for actinide separation, the conversion to chlorides through oxides has been also established through uranium tests.

• -|It is confirmed that more than 99% of TRU nuclides can be recovered from the high level liquid waste by TRU tests

• -|Through these studies, the process flow sheets for reprocessing of metal and oxide fuels and for partitioning of TRU separation have been established.

The subjects to be emphasized for further development are classified into three categories, that is, process development (demonstration), technology for engineering development, and supplemental technology.

The metal fuel FBR has a high potential for recycling actinides by integration with pyro-reprocessing. Alloys of U-Pu-Zr with minor actinides are investigated from points of fuel properties. The miscibility and other characteristics suggest that the maximum content up to ca. 5 wt% of minor actinides is allowable in the matrix. Nine pins of metal fuel including minor actinides are ready for irradiation at Phenix fast reactor.     

亚硝气用于1BP料液的调价

由1BP料液制备2AF料液的过程中需要氧化去除1BP中还原剂,同时将Pu(Ⅲ)氧化至Pu(Ⅳ)。以气液并流模式在玻璃填料柱中研究了亚硝气对模拟1BP料液中还原剂的氧化情形,考察了亚硝气用量、料液停留时间、HNO_3浓度和温度对模拟1BP料液中还原剂氧化率的影响。结果表明:25℃时,在羟胺、肼浓度均约为0.08mol/L,c(HNO_3)=1.5mol/L、ρ_0(U(Ⅳ))=1.26g/L、Pu(Ⅲ)(以Fe(Ⅱ)代替)质量浓度为8.76g/L的1BP料液中,当N_2O_4液体为还原剂总量的1.2倍(摩尔比)、停留时间为3min时,1BP料液中还原剂的氧化率达到99.9%以上,Pu(Ⅲ)的氧化率在99.9%以上。     

Potential of CANDLE reactor on sustainable development and strengthened proliferation resistance

CANDLE reactor generates energy by using only natural or depleted uranium as make up fuel and achieves about 40% burn up without fuel recycling (Sekimoto, H., Ryu, K., Yoshimura, Y., 2001. CANDLE: the new burnup strategy. Nucl. Sci. Eng. 139 (3), 306–317). These distinctive characteristics eliminate the necessity of both enrichment and reprocessing processes that are recognized as essentially inevitable parts in the conventional nuclear energy concept.

This paper describes that the potential performance of CANDLE reactor to meet a projected energy demand growth of the world with stabilizing carbon dioxide concentration in the atmosphere and simultaneously minimize the risk of nuclear material proliferation.

A type of CANDLE reactor with moderate initial fissile inventory is feasible to be deployed with prompt enough introduction pace to satisfy the worldwide energy growth and limit the carbon dioxide concentration to about 550 ppm in the next century. Ultimately high proliferation resistance performance as a fission energy system is found for the CANDLE system due to the elimination of most vulnerable processes in the conventional fuel cycle.     

Neutronics assessment of the use of thorium fuels in current pressurized water reactors

The use of thorium fuel in current PWRs in a once-through fuel cycle is an attractive option due to potential advantages such as high conversion ratio and low minor actinide generation. The current neutronics assessments indicate that the thorium fuel cycle could supplement the current uranium–plutonium fuel cycle to improve operational performance and spent fuel consideration in current PWRs without core and subassembly modifications. Neutronics safety parameters in the PWR cores with the thorium fuels are within the range of current PWRs.The PWR cores with thorium fuels have significantly higher conversion ratios which could enable efficient fuel utilization. Further, it is shown that the use of thorium as a fertile material can reduce minor actinide generation and the radio-toxicity of spent fuels. In considerations related to proliferation resistance, the results of the current analyses show no significant difference between the studied thorium fuels and the standard oxide fuel for the assumed characteristics and burnup levels.     

Impact of Partitioning and Transmutation on High-Level Waste Disposal for the Fast Breeder Reactor Fuel Cycle

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large storage capacity because the number of waste packages produced is significantly reduced by a factor of 5 from that of the glass waste package in the FBR fuel cycle without PT.     

Plutonium management with thorium-based fuels and inert matrix fuels in thermal reactor systems

The plutonium that is produced by light water reactors worldwide is currently re-used to a limited extent. In the last century, the expected introduction of fast reactors and the associated need for large amounts of plutonium did not take place. The result is that worldwide a stockpile of excess plutonium has formed, which is the dominant contributor to the radiotoxicity of spent nuclear fuel for storage times from 10² to 10⁵ years. One option to reduce or stabilize the plutonium stockpile is to utilize this plutonium in advanced fuel types, such as thorium-based and inert matrix fuels. Because these fuels do not contain uranium, the plutonium consumption rate is very high. In this paper, the status of the fuel research and some recent developments are given.