Cascade subcritical molten salt reactor (CSMSR): Main features and restrictions

The ADS-burner of minor actinides (MA) is considered of the following assumptions:

• Proton accelerator — driver has energy E_a=1 GeV and current I_a=10 mA;

• Subcriticality is Δk = 0.05 (k_eff=0.95);

• Reactor consists of two cores: core-1 of cascade amplification (CA) and core-2 of transmutation (CT);

• Molten salt NaF-ZrF₄ is used as a coolant and the nuclear fuel solvent in CT.

Technical solutions are chosen close to used in reactor technology. The main neutron physics characteristics of reactor are calculated including reactor power distribution, reactivity effects, MA burnup, thermo-hydraulics of CA fuel elements etc.Such CSMSR with power 800 MWth can incinerate 50 kg MA/year, i.e. the MA production of five thermal reactors of the same power.     

Study on fuel handling system suitable for MA loading of commercial FBR

On the commercial stage of FBR, the improvement of the fuel handling system which directly removes spent fuels from the reactor without in-vessel storage, will be useful to enhance an effective utilization of nuclear resources.

Minor actinides (MA) loading measures such as the heterogeneous loading (two kinds of MA contents-fuel assemblies are loaded to the same reactor) and the homogeneous loading were extracted, and their suitability for the commercial FBR was estimated from standpoints of the fuel handling system cost and the contribution to the smooth fuel recycle flow.

And it was clarified that the heterogeneous MA loading was useful for the transition period from LWRs to FBRs due to the promotion for the Pu utilization and the MA transmutation, and the homogeneous MA loading for the period of only FBRs due to the reduction for the TRU (Pu and MA) inventory outside the reactor.     

Performance of natural uranium- and thorium-fueled fast breeder reactors (FBRs) for U fissile production

The performance of natural uranium and thorium-fueled fast breeder reactors (FBRs) for producing ²³³U fissile material, which does not exist in nature, is investigated. It is recognized that excess neutrons from FBRs with good neutron economic characteristics can be efficiently used for producing ²³³U. Two distinct metallic fuel pins, one with natural uranium and another with natural thorium, are loaded into a large sodium-cooled FBR. ²³³U and the associated-U isotopes are extracted from the thorium fuel pins. The FBR itself is self-sustained by plutonium produced in the uranium fuel pins. Under the equilibrium state, both uranium and thorium spent fuels are periodically discharged with a certain discharge rate and then separated. All discharged fission products are removed and all discharged actinides are returned to the FBRs except the discharged uranium utilized for fresh fuel of the other thorium-cycled reactors. ²³³U-production rate of the FBRs as a function of both the uranium–thorium fuel pins fraction in the core and the discharge fuel burnup is estimated. The result shows that larger fraction of uranium pins is better for the FBR criticality while larger fraction of thorium fuel pins and lower fuel burnup give higher ²³³U production rate.     

Impact of PWR spent fuel variations on back-end features of advanced fuel cycles with tru-fueled VHTR

Alternative strategies are being considered as management option for current spent nuclear fuel transuranics (TRU) inventory. Creation of transmutation fuels containing TRU for use in thermal and fast reactors is one of the viable strategies. Utilization of these advanced fuels will result in transmutation and incineration of the TRU. The objective of this study is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled very high temperature reactor (VHTR) systems. The current effort is focused on prismatic core configuration operated under a single batch once-through fuel cycle option. IAEA’s nuclear fuel cycle simulation system (VISTA) was used to determine potential PWR spent fuel compositions. Additional composition was determined from the analysis of United States legacy spent fuel that is given in the Yucca Mountain Safety Assessment Report. A detailed whole-core 3-D model of the prismatic VHTR was developed using SCALE5.1 code system. The fuel assembly block model was based on Japan’s HTTR fuel block configuration. To establish a reference reactor system, calculations for LEU-fueled VHTR were performed and the results were used as the basis for comparative studies of the TRU-fueled systems. The LEU fuel is uranium oxide at 15% ²³⁵U enrichment. The results showed that the single-batch core lifetimes ranged between 5 and 7 years for all TRU fuels (3 years in LEU), providing prolonged operation on a single batch fuel loading. Transmutation efficiencies ranged between 19% and 27% for TRU-based fuels (13% in LEU). Total TRU material contents for disposal ranged between 730 and 808 kg per metric ton of initial heavy metal loading, reducing TRU inventory mass by as much as 27%. Decay heat and source terms of the discharged fuel were also calculated as part of the spent fuel disposal consideration. The results indicated strong potential of TRU-based fuel in VHTR.     

Application and evaluation of different numerical methods for determining crack resistance curves

The engineering assessment of precracked components is frequently carried out with the help of crack resistance curves on the basis of the J-integral. Nevertheless, there are severe uncertainties regarding the validity of the J-integral in the case of large plastic deformations and unloading processes due to crack growth. Numerical and theoretical derivations are used to examine the influence of large plastic deformations and stable crack growth on the calculation of the J-integral. Numerical investigations were carried out on the example of a CT 25-specimen made of 20 MnMoNi 5 5 by means of the finite-element method. The following methods of calculations were selected:

• - Calculation of the specimen behaviour without consideration of the stable crack growth.

• - Calculation of the specimen behaviour using a J-Δa-curve as crack crack growth law.

• - Calculation of the specimen behaviour using a damage model (“ocal approach”) to compute the crack growth.

The results of the calculation carried out on the basis of the damage model makes it possible to carry out an assessment of the various methods for the experimental determination of the J-value.     

Electrorefining of metallic fuel with burn-up of ~7 at% in a LiCl-KCl melt

Electrorefining of irradiated metallic fuels (burn-up ~ 7 at%) in a LiCl-KCl melt at 773 K was successfully demonstrated: Actinides in the fuels were anodically dissolved in the melt. Both a selective U metal deposition on a solid cathode and a grouped recovery of actinides, U, Pu, Np, Am, and Cm, in a liquid Cd cathode were confirmed. The behavior of fission products, such as lanthanides, alkali metals, alkaline earth metals, and noble metals, were also investigated. It was found that the behaviors of actinides and fission products in the electrorefining of the fuels with ~ 7 at% burn-up were similar to those in electrorefining of fuels with ~ 2.5 at% burn-up.     

Symbiotic systems consisting of large-FBR and small water-cooled thorium reactors (WTR)

Small long life water-cooled thorium reactors (WTR; 30–300 MWth) have been investigated. For realizing thorium cycle of the reactors, a uranium–thorium mixture core is introduced to fast breeder reactors (FBR; 3000 MWth) to be a ²³³U producer. In the present study, two distinct metallic fuel pins, with natural uranium and thorium, are loaded into a large sodium-cooled FBR. The FBR itself is self-sustained by the plutonium produced in the uranium pins. Under the equilibrium burnup state, the FBR spent fuels are periodically discharged with a certain discharge rate and then separated. Some actinides are returned to the FBR core while ²³³U, which is discharged from the thorium pins, is utilized for the WTR fresh fuel. Fissile support capability is the main investigated parameter of the study. The system achieves higher support capability at higher burnup and lower power of the WTR, and shows that larger number of uranium pins is better for the FBR criticality while larger number of thorium pins and lower burnup give better support factor capability. For a symbiotic system consisting 3000 MWth FBR and 100 MWth WTRs, where discharged fuel burnup is 96 and 60 GWd/t for the FBR and WTRs, one FBR can support 5 WTRs.     

Assessment of criticality and burn up behavior of candu reactors with nuclear waste trans uranium fuel

Large quantities of nuclear waste plutonium and minor actinides (MAs) have been accumulated in the civilian light water reactors (LWRs) and CANDU reactors. These trans uranium (TRU) elements are all fissionable, and thus can be considered as fissile fuel materials in form of mixed fuel with thorium or nat-uranium in the latter. CANDU fuel compacts made of tristructural-isotropic (TRISO) type pellets would withstand very high burn ups without fuel change.As carbide fuels allow higher fissile material density than oxide fuels, following fuel compositions have been selected for investigations: ① 90% nat-UC + 10% TRUC, ② 70% nat-UC + 30% TRUC and ③ 50% nat-UC + 50% TRUC. Higher TRUC charge leads to longer power plant operation periods without fuel change. The behavior of the criticality k_∞ and the burn up values of the reactor have been pursued by full power operation for > ∼12 years. For these selected fuel compositions, the reactor criticality starts by k_∞ = 1.4443, 1.4872 and 1.5238, where corresponding reactor operation times and burn up values have been calculated as 2.8 years, 8 years and 12.5 years, and 62, 430 MW.D/MT, 176,000 and 280,000 MW.D/MT, with fuel consumption rates of ∼16, 5.68 and 3.57 g/MW.D respectively. These high burn ups would reduce the nuclear waste mass per unit energy output drastically. The study has show clearly that TRU in form of TRISO fuel pellets will provide sufficient criticality as well as reasonable burn up for CANDU reactors in order to justify their consideration as alternative fuel.     

A new concept of nuclear fuel reprocessing by applying ion-exchange technology

The development of advanced technology for the spent nuclear fuel reprocessing should be achieved not only considering cost, non proliferation and reduction of radioactive wastes but also corresponding to both spent nuclear fuels of LWR and FBR.

We have proposed an ion exchange process for reprocessing using a new type ion exchanger developed to chemical method of U enrichment technology. This process possess possibility of a sharp cut in cost, since this ion exchanger is characterized by rapid adsorption-desorption rate dominating the treatment rate.

From the basic experimental results, this reprocessing process has been constructed by 3 ion exchanger columns which consist of a main separation column, the uranium-refining column and the plutonium-refining column.

Comparing ion exchange process with the conventional Purex process, this ion exchange process has many advantages such as the decrease in the number and size separation equipment, solvent-spent free and alkaline-liquid-spent free. With these advantages, this process is estimated that the construction cost of reprocessing process is greatly reduced comparing to the conventional process.     

Transmutation of minor actinides and innovative fuel cycle concepts

Both physical and chemical properties of actinides show significant variations depending on the element from uranium to curium. The materials database on minor actinides (MA:Np, Am, Cm) is very limited. Even with the scanty database, however, americium and curium did not appear to make viable fuels in the ordinary sense. The solutions may be (1) to tolerate a significantly lower fuel performance whose penalty has to be recovered by an improved overall economy of the MA-fuel cycle, and/or (2) to make a fuel concept whose performance is less restricted by physical and chemical properties of MA. In such a difficult program like actinide burning, the technology components would have to be combined by a modular approach. The system has to be integrated as a whole, but a technological module in the system has to be made as such that can be replaced by the other modules. This type of approach would be only possible if we have a technology which forms a common basis to these “modules”. The advantage of taking the modular approach is to allow the system to evolve with time. The pyrochemical separation can form such a common basis.     

Decay heat production in a TRU burner

The design of new reactors such as ADS has been investigated in many countries during the last years for burning transuranic nuclides (TRUs) contained in spent reactor fuel. To increase the TRU incineration rate, fertile-free dedicated fuels, which may contain a large fraction of minor actinides (MAs), are currently considered. Based on past experience, R&D activities for dedicated fuels in Europe concentrate on fuel forms, in which the oxide actinide phase is mixed with oxide or metal inert matrices. Decay heat in a system with inert matrix fuel (IMF) containing MAs may differ from that in a conventional fast reactor. In this paper, several fast reactor designs with different TRU content are considered and related decay heat values, calculated on the basis JEFF 3.0 and JEFF 3.1 nuclear data libraries, are compared. It is shown that some decay heat components for fuels with MAs may be lower than those for MA-free fuels, but the total decay heat may be significantly higher for cooling times exceeding about 1 min.     

Scenarios for the transmutation of actinides in CANDU reactors

With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU(R) reactor.Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio.This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past ( [Boczar et al., 1996] , [Chan et al., 1997] , [Hyland and Dyck, 2007] and [Hyland et al., 2009] ). The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel.The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100–1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.     

Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel

Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U–Zr alloy fuel elements irradiated in the Experimental Breeder Reactor II (EBR-II). We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments (hulls) that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr (and U) found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining.     

BERN railway station: security assessment under a speed increase of trains from 30 to 40 km h

In the frame of the “Rail 2000” project, the CFF (Swiss Railways) would like to increase the speed of trains arriving in Bern station. On the eastern head, this speed would be raised from 30 to 40 km h⁻¹. The superstructure is formed by a three-storey building which rests on 450 mm diameter steel columns.

The aim of the present study was:

• - to determine the security loss of the station superstructure under a train impact on the columns at 40 km h⁻¹ rather than at 30 km h⁻¹

• - to propose measures in order to get at 40 km h⁻¹ the same security as at 30 km h⁻¹

Four approaches are dealt with:

1. (1) on the base of accidents statistics and of their cost;

2. (2) review of possible dynamical approaches;

3. (3) equivalent static load (from European railways codes);

4. (4) energy in which, starting from its initial speed, the train loses energy on different obstacles (ballast, platform, protection devices, walls, train's own deformation) and the remaining energy is compared with the maximum energy that the column can dissipate by deformation.

The conclusions are presented as a ‘security plan’ and the proposed protection costs are evaluated.     

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.     

A blueprint for GNEP advanced burner reactor startup fuel fabrication facility

The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu)-239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10-15 years and would take 3-5 years to construct.     

Modeled salt density for nuclear material estimation in the treatment of spent nuclear fuel

Spent metallic nuclear fuel is being treated in a pyrometallurgical process that includes electrorefining the uranium metal in molten eutectic LiCl-KCl as the supporting electrolyte. We report a model for determining the density of the molten salt. Material balances account for the net mass of salt and for the mass of actinides present. It was necessary to know the molten salt density, but difficult to measure. It was also decided to model the salt density for the initial treatment operations. The model assumes that volumes are additive for the ideal molten salt solution as a starting point; subsequently, a correction factor for the lanthanides and actinides was developed. After applying the correction factor, the percent difference between the net salt mass in the electrorefiner and the resulting modeled salt mass decreased from more than 4.0% to approximately 0.1%. As a result, there is no need to measure the salt density at 500 °C for inventory operations; the model for the salt density is found to be accurate.     

The separation of fission-product rare elements toward bridging the nuclear and soft energy systems

Based on the present state of the art of the separation technology, recycling of fission-product rare elements (FRE) in the FBR spent fuel is discussed. The rad.-waste fractionation is in accordance with the present society's trend toward zero-emission, and the mean of salt-free method utilizing electrochemistry agrees with the principles of the newly established green chemistry. A catalytic electrolytic extraction method is proposed to separate the target, radioactive but potentially strategic elements, Pd, Ru, Rh, Re (Tc), Te and Se dissolved in the HLLW. It avoids secondary waste arising. This method is particularly feasible for the separation of Pd where cyclic reaction of metal cations such as Pd(II) or Fe(II), acting as promoters or mediators and already contained in HLLW, accelerates the electrochemical deposition of Ru, Rh and Re. Elemental separation not only offers alternative material resources to meet expanding demands for catalysts in Fuel Cell/Soft Energy system but is also the first step for transmutation or other selective strategies for waste management of long-lived fission products (LLFP).     

Advanced fuel cycle scenarios for High Temperature Gas Reactors

This study evaluates nuclear fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. Three fuel cycle scenarios, the Light Water Reactor (LWR)–Advanced Recycling Reactor (ARR) recycle, the LWR–High Temperature Gas Reactor (HTGR)–ARR recycle, and the HTGR partial recycling fuel cycle, are assessed for their mass flow and electricity generation costs and the results are compared to those of the LWR once-through fuel cycle. The spent fuels are recycled in both the Consolidated Fuel Treatment Center and the Actinide Management Island, which are capable of reprocessing spent fuels by Uranium Extraction and Pyrochemical processes, respectively. The mass flow calculations show that the Transuranics (TRU) which have a long-term radiation effect can be completely burned in the recycling fuel cycles, resulting in 350, 450 and 6 times reduction of TRU inventory for the LWR–ARR, LWR–HTGR–ARR and HTGR partial recycling fuel cycles, respectively, when compared to the once-through fuel cycle. The electricity generation costs of these fuel cycle scenarios were estimated to be 39.1, 34.9 and 25.7 USD/MW h(e), which are comparable to or smaller than that of the once-through fuel cycle. Although the candidate fuel cycles adopt reprocessing options which raise fuel cycle cost, increase in uranium cost and the advanced design of the HTGR can further reduce the advanced fuel cycle costs of the HTGR.     

Electrolytic Reduction of Spent Light Water Reactor Fuel Bench-Scale Experiment Results

A series of experiments were performed to demonstrate the electrolytic reduction of spent light water reactor fuel at bench-scale in a hot cell at the Idaho National Laboratory Materials and Fuels Complex. The process involves the conversion of oxide fuel to metal by electrolytic means, which would then enable subsequent separation and recovery of actinides via existing electrometallurgical technologies, i.e., electrorefining. Four electrolytic reduction runs were performed at bench scale using ～500 ml of molten LiCl–1 wt% Li₂O electrolyte at 650°C. In each run, ～50 g of crushed spent oxide fuel was loaded into a permeable stainless steel basket and immersed into the electrolyte as the cathode. A spiral wound platinumwire was immersed into the electrolyte as the anode. When a controlled electric current was conducted through the anode and cathode, the oxide fuel was reduced to metal in the basket and oxygen gas was evolved at the anode. Salt samples were extracted before and after each electrolytic reduction run and analyzed for fuel and fission product constituents. The fuel baskets following each run were sectioned and the fuel was sampled, revealing an extent of uranium oxide reduction in excess of 98%.