Reactivity feedback calculation of a conceptual TRISO fueled compact PWR core

Reactivity feedback coefficients have been calculated for a compact sized PWR core that utilizes carbon coated micro fuel particles instead of standard cylindrical fuel pellets with an inventive composition. A small amount of Pu-240 with 5 w/o has also been added in tristructural-isotropic (TRISO) fuel in place of U-238 for the reduction of excess reactivity. The values of fuel, moderator and void reactivity coefficients have been calculated at the middle of fuel cycle. All the reactivity coefficients were found negative which meet the design safety criteria. It was also observed that all reactivity feedback coefficients are interlinked and their effects are pronounced when coupled together.     

Characteristics of a different fuel cycle in a PBMR-400 for burning reactor grade plutonium

An outstanding feature of the high temperature, gas-cooled, version of a Generation IV type reactor is its versatility in application. Apart from the capacity in high temperature gas-cooled reactors to generate electricity it could desalinate (multi-effect distillation [MED] by deploying excess heat or reverse osmosis by deploying excess electricity), produce hydrogen (by deploying excess electricity), whereas this article showcases the on-line fuelling characteristics of a pebble bed reactor concept for the incineration of reactor grade plutonium, whilst producing electricity.The VSOP-A system of codes is employed to demonstrate by calculation how the standard PBMR-400 commercial reactor design offers similar inherent safety characteristics with a Pu-Th/U advanced fuel cycle. This implies that no significant design changes are necessary to implement such a fuel cycle. Furthermore, the flexibility of the pebble fuel concept is deployed to house the fertile material in one type of pebble, whilst a second type will contain the fissile or driver material.     

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.     

Criticality and burnup analyses of a PBMR-400 full core using Monte Carlo calculation method

In this study, the criticality and burnup analyses have been performed for full core model of Pebble Bed Modular Reactors, such as PBMR-400, using the computer codes MCNP5.1.4 and MONTEBURNS 2.0. Three different pebble distributions, namely; Body Centered Cubic (BCC) (packing fraction = 68%), Random Packing (RP) (packing fraction = 61%) and Simple Cubic (SC) (packing fraction = 52%) were selected for the analyses. The calculated core effective multiplication factor, k_eff, for BCC, RP and SC came to be 1.2395, 1.2357 and 1.2223, respectively. The core life for these distributions were calculated as ～1200, 1000, and 800 Effective Full Power Days (EFPDs), whereas, the corresponding burnups came out to be ～99,000, ～92,000 and ～86,000 MWD/T, respectively, for end of life k_eff set equal to 1.02.     

Effect of minor actinides doping on plutonium produced in large-scale FBR blanket

The present study focuses on the effect of minor actinides (MAs) addition into the FBR blanket as ways of increasing fraction of even-mass-number plutonium isotopes, especially ²³⁸Pu, aiming at enhancing the proliferation resistance of plutonium produced in the blanket. The MA loading potential to enhance the proliferation resistance of plutonium is investigated, with considering actual design constraints on the fuel decay heat from the fuel handling and fabrication points of view, as MAs considerably generate decay heat. It reveals that depending on doping quantity of MAs, it is possible to denature produced plutonium by MA transmutation. MA addition in the blanket gives a significant increment in ²³⁸Pu fraction of generated plutonium but less effect on other even-mass-number plutonium isotopes. However, it is important that MA compositions should be adequately controlled to satisfy both the proliferation resistance requirements and the decay heat constraints for fuel handling.     

Influence of void fraction change on plutonium and minor actinides recycling in BWR with equilibrium burnup

A study on the influence of void fraction change on plutonium and minor actinides recycling in standard boiling water reactor (BWR) with equilibrium burnup model has been conducted. We considered the equilibrium burnup model since it is a simple time independent burnup method that can handle all possible produced nuclides in any nuclear system.

The uranium enrichment for the criticality of the reactor diminishes significantly for the plutonium and minor actinides recycling case compared to that of the once-through cycle of BWR case. This parameter decreases much lower with the increasing of the void fraction. A similar propensity was also shown in the required natural uranium per annum. The annual required natural uranium was calculated by assuming that the uranium concentration in the tail of the enrichment plant is 0.25 w%. The amount of loaded fuel reduces slightly with the increment of the void fraction for plutonium and minor recycling in BWR.     

Utilization of TRISO fuel with reactor grade plutonium in CANDU reactors

Large quantities of plutonium have been accumulated in the nuclear waste of civilian LWRs and CANDU reactors. Reactor grade plutonium and heavy water moderator can give a good combination with respect to neutron economy. On the other hand, TRISO type fuel can withstand very high fuel burn-up levels. The paper investigates the prospects of utilization of TRISO fuel made of reactor grade plutonium in CANDU reactors. TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The fuel compacts conform to the dimensions of CANDU fuel compacts are inserted in rods with zircolay cladding.In the first phase of investigations, five new mixed fuel have been selected for CANDU reactors composed of 4% RG-PuO₂ + 96% ThO₂; 6% RG-PuO₂ + 94% ThO₂; 10% RG-PuO₂ + 90% ThO₂; 20% RG-PuO₂ + 80% ThO₂; 30% RG-PuO₂ + 70% ThO₂. Initial reactor criticality (k_∞,0 values) for the modes , , , and are calculated as 1.4294, 1.5035, 1.5678, 1.6249, and 1.6535, respectively. Corresponding operation lifetimes are ∼0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ∼30 000, 60 000, 100 000, 200 000 and 290 000 MW d/tonne, respectively. The higher initial plutonium charge is the higher burn ups can be achieved.In the second phase, a graphical-numerical power flattening procedure has been applied with radially variable mixed fuel composition in the fuel bundle. Mixed fuel fractions leading to quasi-constant power production are found in the 1st, 2nd, 3rd and 4th row to be as 100% PuO₂, 80/20% PuO₂/ThO₂, 60/40% PuO₂/ThO₂, and 40/60% PuO₂/ThO₂, respectively. Higher plutonium amount in the flattened case increases reactor operation lifetime to >8 years and the burn up to 580 000 MW d/tonne.Power flattening in the bundle leads to higher power plant factor and quasi-uniform fuel utilization, reduces thermal and material stresses, and avoids local thermal peaks. Extended burn-up grade implies drastic reduction of the nuclear waste material per unit energy output for final waste disposal.     

Criticality investigations for the fixed bed nuclear reactor using thorium fuel mixed with plutonium or minor actinides

Prospective fuels for a new reactor type, the so called fixed bed nuclear reactor (FBNR) are investigated with respect to reactor criticality. These are ① low enriched uranium (LEU); ② weapon grade plutonium + ThO₂; ③ reactor grade plutonium + ThO₂; and ④ minor actinides in the spent fuel of light water reactors (LWRs) + ThO₂. Reactor grade plutonium and minor actinides are considered as highly radio-active and radio-toxic nuclear waste products so that one can expect that they will have negative fuel costs.The criticality calculations are conducted with SCALE5.1 using S₈–P₃ approximation in 238 neutron energy groups with 90 groups in thermal energy region. The study has shown that the reactor criticality has lower values with uranium fuel and increases passing to minor actinides, reactor grade plutonium and weapon grade plutonium.Using LEU, an enrichment grade of 9% has resulted with k_eff = 1.2744. Mixed fuel with weapon grade plutonium made of 20% PuO₂ + 80% ThO₂ yields k_eff = 1.2864. Whereas a mixed fuel with reactor grade plutonium made of 35% PuO₂ + 65% ThO₂ brings it to k_eff = 1.267. Even the very hazardous nuclear waste of LWRs, namely minor actinides turn out to be high quality nuclear fuel due to the excellent neutron economy of FBNR. A relatively high reactor criticality of k_eff = 1.2673 is achieved by 50% MAO₂ + 50% ThO₂.The hazardous actinide nuclear waste products can be transmuted and utilized as fuel in situ. A further output of the study is the possibility of using thorium as breeding material in combination with these new alternative fuels.     

Neutronic and safety aspects of inert matrix fuel utilization in fast reactors for plutonium and minor actinides transmutation

The solving of ecological problems of future nuclear power is connected with the solving of long-lived radioactive waste utilization problems. It concerns primarily plutonium and minor actinides (MAs), accumulated in the spent fuel of nuclear reactors. One of the ways this can be solved is to use a fast reactor with uranium-free or inert matrix fuel (IMF). The physics of this type of reactor was widely investigated during last year for BN-800 reactors. The solution of the most important problems was: a decrease in non-uniformity of power distribution and an increase of the Doppler effect. The next stage of such core investigations is an evaluation of self-protection to beyond design accidents. Preliminary results show a high safety level of BN-800 reactors with IMF in the event of unprotected loss of coolant flow (ULOF) accident.     

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.     

Neutronics analysis of minor actinides transmutation in a fusion-driven subcritical system

The minor actinides (MAs) transmutation in a fusion-driven subcritical system is analyzed in this paper. The subcritical reactor is driven by a tokamak D-T fusion device with relatively easily achieved plasma parameters and tokamak technologies. The MAs discharged from the light water reactor (LWR) are loaded in transmutation zone. Sodium is used as the coolant. The mass percentage of the reprocessed plutonium (Pu) in the fuel is raised from 0 to 48% and stepped by 12% to determine its effect on the MAs transmutation. The lesser the Pu is loaded, the larger the MAs transmutation rate is, but the smaller the energy multiplication factor is. The neutronics analysis of two loading patterns is performed and compared. The loading pattern where the mass percentage of Pu in two regions is 15% and 32.9% respectively is conducive to the improvement of the transmutation fraction within the limits of burn-up. The final transmutation fraction of MAs can reach 17.8% after five years of irradiation. The multiple recycling is investigated. The transmutation fraction of MAs can reach about 61.8% after six times of recycling, and goes up to about 86.5% after 25.     

Evaluation on transmutation performance of minor actinides with high-flux BWR

The performance of high-flux BWR (HFBWR) for burning and/or transmutation (B/T) treatment of minor actinides (MA) and long-lived fission products (LLFP) was discussed herein for estimating an advanced waste disposal with partitioning and transmutation (P&T). The concept of high-flux B/T reactor was based on a current 33 GWt-BWR, to transmute the mass of long-lived transuranium (TRU) to short-lived fission products (SLFP). The nuclide selected for B/T treatment was MA (Np-237, Am-241, and Am-243) included in the discharged fuel of LWR. The performance of B/T treatment of MA was evaluated by a new function, i.e. [F/T ratio], defined by the ratio of the fission rate to the transmutation rate in the core, at an arbitrary burn-up, due to all MA nuclides. According to the results, HFBWR could burn and/or transmute MA nuclides with higher fission rate than BWR, but the fission rate did not increase proportionally to the flux increment, due to the higher rate of neutron adsorption. The higher B/T fraction of MA would result in the higher B/T capacity, and will reduce the units of HFBWR needed for the treatment of a constant mass of MA. In addition, HFBWR had a merit of higher mass transmutation compared to the reference BWR, under the same mass loading of MA.     

Design and optimization of an equilibrium reload with MOX fuel with minor actinides

In this work the design and optimization of an equilibrium core for a boiling water reactor (BWR), loaded with fuel composed of plutonium and minor actinides (Np, Am and Cm), is presented. The plutonium and minor actinides are obtained from the recycling of the spent fuel of a BWR, and are mixed with depleted uranium obtained from the enrichment tails. The design and optimization of the equilibrium reload is achieved in two steps. In the first step, the fuel assembly is adjusted and the reload pattern is designed, in order to obtain the target cycle length. In order to improve the shutdown margin, two actions were taken: to increase the boron-10 content in the control rods, and to add a burnable absorber (gadolinia) in some fuel rods. In the second step, the reload pattern, obtained in the first step, is optimized to maximize the energy, under the thermal and reactivity margins constraints; a system based on Genetic Algorithms was used in the optimization process. Results show that 5% more energy was obtained with the optimized reload.     

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.     

Migration of minor actinides and lanthanides in fast reactor metallic fuel

Minor actinides (MA) and lanthanides (LA) migration to the cladding in fast reactor metallic fuel is a concern because of their reaction with the cladding material. MA and LA migration test results are analyzed and reactions with the cladding are characterized. Remedies for reducing MA and LA migration and reaction with cladding are reviewed. A possible method is proposed that includes the addition of a compound forming element such as indium or thallium in fuel. These elements preferentially form stable compounds with MA and LA and should therefore reduce migration of MA and LA. As an intrinsic solution to the issue, the feasibility of the use of U-Pu-Mo alloy as fuel is studied. Unlike U-Pu-Zr, U-Pu-Mo consists of a single phase at typical fuel operation temperatures, and should have negligible fuel constituent redistribution and reduce MA and LA migration.     

The treatment of TRISO coated particles with CF4 in a low temperature plasma

An alternative recovery method to the mechanical crushing of off-specification tri-structural-isotropic (TRISO) coated fuel microspheres is demonstrated. It is shown that the inert SiC layer can be completely removed by etching with the active fluorine species from an inductively coupled radio-frequency CF₄ glow-discharge impinging a static bed from the top, at a working pressure of 1 kPa. At this pressure mass transport does not have a rate limiting role and the chemical reaction itself is rate determining. A treatment time of roughly 4 h is required for the conditions reported here.     

Transmutation characteristics of minor actinides in a low aspect ratio tokamak fusion reactor

The transmutation characteristics of minor actinides in the transmutation reactor of a low aspect ratio (LAR) tokamak are investigated. One-dimensional neutron transport and burn-up calculations coupled with a tokamak systems analysis were performed to determine optimal system parameters. The dependence of the transmutation characteristics, including the neutron multiplication factor, produced power, and the transmutation rate, on the aspect ratio A in the range of 1.5–2.0 was examined. By adding Pu239 to the transmutation blanket as a neutron multiplication material, it was shown that a single transmutation reactor producing a fusion power of 150 MW_th can destroy minor actinides contained in the spent fuels for more than 38 units of 1 GW_e pressurized water reactors (PWRs) while producing a power in the range of 1.8–6.8 GW_th.     

The GDT-based fusion neutron source as driver of a minor actinides burner

At present, accelerator driven systems seem to have a good chance to play an important role in a long-term sustainable utilization of fission reactor technology. Compared with reactors, such systems offer several advantages regarding the incineration of minor actinides. However, they demand intense neutron sources. The spallation neutron source is favoured for this purpose because of its high neutron emission intensity, which seems to be attainable. The Budker Institute of Nuclear Physics Novosibirsk is developing a 14 MeV fusion neutron source on the base of the gas dynamic trap (GDT), which is primarily destined for an irradiation facility for fusion material research. The potential of this neutron source as driver of a minor actinides burner was studied by means of neutron transport calculations and compared with a spallation source. To this end, a simple model of a burner was derived from an international numerical benchmark exercise that was conducted by the Nuclear Energy Agency of the OECD. The paper presents and discusses the main results of the study and draws the conclusion that both the source strength and the efficiency of the GDT-based neutron source must be substantially increased. Moreover, advices are derived, which show that by stretching the neutron production volume and by raising the electron temperature of the GDT plasma the desired improvements could be accomplished.     

Extraction separation of trivalent minor actinides from lanthanides with hydrophobic derivatives of TPEN

We are developing a new MA/Ln separation process with TPEN (N,N,N′,N′-tetrakis(2-pyridylmethyl)ethylenediamine) and its derivatives for P&T technology of HLW from spent nuclear fuel reprocessing plants. TPEN is a hexadentate ligand that has six soft-donor sites as a kind of podand type molecule and can encapsulate a metal ion. TPEN has good selectivity of Am(III) from Ln(III) and has potential to establish partitioning of MA. However, there is a serious problem for the practical application. This is to the dissolution of a slight amount of TPEN (about 10⁻⁴ mol/L) to water. High enrichment of Am(III) will be restricted by the dissolution of TPEN to water. In this study, the hydrophobicity of TPEN is improved by introducing alkyl groups and the effect of the introduction of alkyl groups on the separation of Am(III) and Eu(III) is examined. We tried to synthesize four hydrophobic derivatives of TPEN, and three derivatives were synthesized successfully. The derivatives were examined both the extractability and selectivity of Am(III) and Eu(III). One of them, tpdben, showed good selectivity and the maximum separation factor, SF_Am/Eu, was 34 at pH 5.06. A hydrophobic derivative of TPEN that has potential of application to the MA/Ln separation process was synthesized successfully.     

Optimization of flux distribution in a molten-salt reactor with a 2-region core for plutonium burning

Molten-salt reactors (MSRs) are selected as one of the candidates of Generation IV reactor concepts. In GLOBAL2005 held in Tsukuba, Japan, one paper discussed the flattening of fast neutron flux in the core for a longer life of graphite moderator. In the paper a 3-region reactor concept was presented. The authors tried many cores changing configurations such as volume of each region and fractions of fuel salt in the regions or fuel compositions.

We investigated the other possibility of a 2-region core for the simplicity. Using one energy group neutron diffusion theory and considering extrapolation distance, the optimum selection of region wise neutron multiplication factors can be theoretically and easily obtained. In MSRs, there is no burnup distribution of the fuel. The region wise neutron multiplications can be obtained by adjusting the volume fraction of fuel in a cell with a given composition of the fuel salt. Using the theoretical results, the optimization of the actual core configuration was determined by a nuclear analysis code SRAC2002 with the nuclear data library of JENDL3.3.

In this paper, we considered MSRs using plutonium as a fissile material. Ordinary MSRs use uranium-233, which doesn't exist naturally, and utilizing plutonium is easier to startup.