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.     

Increased fuel burn up in a CANDU thorium reactor using weapon grade plutonium

Weapon grade plutonium is used as a booster fissile fuel material in the form of mixed ThO₂/PuO₂ fuel in a Canada Deuterium Uranium (CANDU) fuel bundle in order to assure the initial criticality at startup.Two different fuel compositions have been used: (1) 97% thoria (ThO₂) + 3%PuO₂ and (2) 92% ThO₂ + 5% UO₂ + 3% PuO₂. The latter is used to denaturize the new ²³³U fuel with ²³⁸U. The temporal variation of the criticality k_∞ and the burn-up values of the reactor have been calculated by full power operation for a period of 20 years. The criticality starts by k_∞ = 1.48 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout. The criticality becomes quasi constant after the second year and remains above k_∞ > 1.06 for 20 years. After the second year, the CANDU reactor begins to operate practically as a thorium burner.Very high burn up could be achieved with the same fuel material (up to 500,000 MW·D/T), provided that the fuel rod claddings would be replaced periodically (after every 50,000 or 100,000 MW·D/T). The reactor criticality will be sufficient until a great fraction of the thorium fuel is burnt up. This would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically.     

Evolution of defects in micropellet fuel during the fabrication of HTGR compacts

The stress distribution arising in micropellets and cylindrical fuel compacts during fabrication, the stress concentration in micropellets located near the surface of a compact, and the evolution of defects in micropellets as a function of the type of stress state are investigated.It has been found that an ensemble of micropellets with a large number of particles contains a continuous spectrum of defects in the range 10^–4–10² µm. Mechanical stresses engender evolution of the defects according to the scheme accumulation of microdefects microcracks cracks through defects.Recommendations are formulated for lowering the number of defects in micropellets during deposition of coatings on the micropellets and compaction.Research Institute of the Luch Scientific and Industrial Association.__________Translated from Atomnaya Énergiya, Vol. 98, No. 1, pp. 44–50, January, 2005.     

Performance analysis of coated plutonia particle fuel compact for radioisotope heater units

Coated plutonia particle fuel has been proposed recently for use in radioisotope power systems and radioisotope heater units for a variety of space missions requiring power levels from milliwatts to tens or even hundreds of watts. The ²³⁸PuO₂ fuel kernels are coated with a strong layer of ZrC designed to fully retain the helium gas generated by the radioactive decay of ²³⁸Pu. A recent investigation has concluded that helium retention in large-grain (200 μm) granular and polycrystalline fuel kernels is possible even at high-temperatures (>1700 K). Results of performance analysis showed that this fuel form could increase by 2.3–2.4 times the thermal power output of a light weight radioisotope heater unit. These figures are for a single-size (500 μm) particles compact, assuming 10% and 5% helium gas release respectively, and a fuel temperature of 1723 K, following 10 years of storage. A binary-size (300 and 1200 μm) particles compact increases the thermal power output of the RHU by an additional 15%.     

Criticality theory for a reactor in the case of fluctuations in the fuel density

Conclusions In Fig. 2 we show graphs of the dependence of the additional reactivity that arises as a result of fluctuations of the fuel density. As follows from Fig. 2, the increase in the reactivity for sufficiently large reactors and for (/₀) 0.1–0.05 is comparable with the contribution of the delayed neutrons. Thus, it is in principle possible to regulate the criticality of the reactor by exciting fluctuations of the density of a gaseous fuel. Regulation in this manner has decided advantages. Thus, the time in which the reactivity changes which determines the transient processes, is short — of the order of one period of the fluctuation. Moreover, there is practically no danger of accidents, since the reactivity falls as soon as the fluctuations cease. The amplitude of the fluctuation of the neutron flux (see, for example, the expression (20)) always exceeds the amplitude of the fluctuations in the fuel density by (k–1)^–1/2. This circumstance may be exploited to obtain a neutron flux pulsating with a large amplitude.This effect of a growth in thereactivity as a result of fluctuations of the fuel density may prove important in the study of the possibility of self-oscillatory conditions of operation in reactors with a high neutron flux.Translated from Atomnaya Énergiya, Vol. 27, No. 2, pp. 107–111, August, 1969.     

Thermal hydraulic study on detection of random failure of fuel by delayed neutron detection system

During a fuel element failure in a liquid metal cooled fast breeder reactor, the fission products originating from the failed pins mix into the sodium pool. Delayed neutron detectors are provided in the sodium pool to detect such failures by way of detection of delayed neutrons emitted by delayed neutron precursors present in the fission products. The transient evolution of precursor concentration is governed by the sodium flow distribution in the pool. Transient thermal hydraulic analysis of precursor concentration evolution in the hot pool of a 500 MWe fast reactor has been carried out using the computational fluid dynamics code PHOENICS to estimate the time at which the failures of various fuel subassembly are detected. Standard k– turbulence model is used to take care of turbulence in conjunction with the precursor concentration equation. It has been found that in order to effectively detect the failure of all fuel SA in the core, a minimum of eight detectors are essential to be provided in the hot pool.     

H2 inhibition of radiation induced dissolution of spent nuclear fuel

In order to elucidate the effect of noble metal clusters in spent nuclear fuel on the kinetics of radiation induced spent fuel dissolution we have used Pd particle doped UO₂ pellets. The catalytic effect of Pd particles on the kinetics of radiation induced dissolution of UO₂ during γ-irradiation in containing solutions purged with N₂ and H₂ was studied in this work. Four pellets with Pd concentrations of 0%, 0.1%, 1% and 3% were produced to mimic spent nuclear fuel. The pellets were placed in 10 mM aqueous solutions and γ-irradiated, and the dissolution of was measured spectrophotometrically as a function of time. Under N₂ atmosphere, 3% Pd prevent the dissolution of uranium by reduction with the radiolytically produced H₂, while the other pellets show a rate of dissolution of around 1.6 × 10⁻⁹ mol m⁻² s⁻¹. Under H₂ atmosphere already 0.1% Pd effectively prevents the dissolution of uranium, while the rate of dissolution for the pellet without Pd is 1.4 × 10⁻⁹ mol m⁻² s⁻¹. It is also shown in experiments without radiation in aqueous solutions containing H₂O₂ and O₂ that ?-particles catalyze the oxidation of the UO₂ matrix by these molecular oxidants, and that the kinetics of the catalyzed reactions is close to diffusion controlled.     

Plutonium charging of the VVÉR-440 water-cooled water-moderated reactor

Conclusions Our study has been limited to a neutron-physical analysis of the active zone in the presence of a fresh charge. Allowing for the disposition of the fuel assemblies this reflects a reasonably practical situation. The difficulties associated with local peaks of energy evolution may be quite easily overcome by using different fuel enrichments inside the assembly.Our example of the charging of the active zone with plutonium is of a hypothetical character, since there is far more plutonium in this arrangement than is required in any practical situation. The calculations show that there are no problems as regards the distribution of heat evolution with respect to the radius of the active zone.It would be extremely desirable to pursue this investigation with due allowance for the factors involved in the burnup of the nuclear fuel. Certain experimental work will be required in this connection so as to provide confirmation of the validity of the computer calculations. The nonuniformity of the distribution of energy evolution will be smoothed as the nuclear fuel is impoverished during burnup.Another important aspect to be studied is that of determining the weight of the control rods. We feel that a study of reactivity problems will reveal some more rigorous limitations than those deduced from the study of energy distribution undertaken in the present investigation.State Technical Scientific-Research Center. Laboratory of Nuclear-Power Technology, Helsinki, Finland. Translated from Atomnaya Énergiya, Vol. 40, No. 4, pp. 283–286, April, 1976.     

Results of AVR fuel pebble irradiation at increased temperature and burn-up in the HFR Petten

The irradiation experiment HFR-EU1bis was performed by the European Commission's Joint Research Centre-Institute for Energy (JRC-IE) in the HFR Petten to test five spherical High Temperature Reactor (HTR) fuel pebbles of former German production with TRISO coated particles for their potential for very high temperature performance and high burn-up. The irradiation started on 9 September 2004 and was terminated on 18 October 2005 after 10 reactor cycles totaling 249 efpd and a maximum burn-up of 11.07% FIMA.The objective of the HFR-EU1bis test was to irradiate five HTR fuel pebbles at conditions beyond the characteristics of current HTR reactor designs with pebble bed cores, e.g. HTR-Modul, HTR-10 and PMBR. This should demonstrate that pebble bed HTRs are capable of enhanced performance in terms of sustainability (further increased power conversion efficiency, better use of fuel) and thus reduced waste production. The central temperature of all pebbles was kept as closely as possible at 1250 °C and held constant during the entire irradiation, with the exception of HFR downtime and power transients. This is the expected maximum central fuel temperature of a pebble bed VHTR with a coolant outlet temperature of 1000 °C.HFR-EU1bis should demonstrate the feasibility of low coated particle failure fractions under normal operating conditions and more specifically:

•

increased central fuel temperature of 1250 °C compared to 1000-1200 °C in earlier irradiation tests;

•

irradiation to a burn-up close to 16% FIMA, which is double the license limit of the HTR-Modul; due to a neutronics data processing error, the experiment was prematurely terminated at 11.07% FIMA maximum so that this objective was not fully achieved;

•

confirmation of low coated particle failure fractions due to temperature, burn-up and neutron fluence.

This paper provides the irradiation history of the experiment including data on fission gas release. Post-irradiation examinations at NRG Petten and JRC-ITU Karlsruhe included the verification of the received neutron fluences, burn-up and spectrum. They will be followed shortly by safety-relevant heating tests at JRC-ITU to verify fission product retention by out-of-pile heating tests beyond 1600 °C.     

Optimization of a radially cooled pebble bed reactor

By altering the coolant flow direction in a pebble bed reactor from axial to radial, the pressure drop can be reduced tremendously. In this case the coolant flows from the outer reflector through the pebble bed and finally to flow paths in the inner reflector. As a consequence, the fuel temperatures are elevated due to the reduced heat transfer of the coolant. However, the power profile and pebble size in a radially cooled pebble bed reactor can be optimized to achieve lower fuel temperatures than current axially cooled designs, while the low pressure drop can be maintained.The radial power profile in the core can be altered by adopting multi-pass fuel management using several radial fuel zones in the core. The optimal power profile yielding a flat temperature profile is derived analytically and is approximated by radial fuel zoning. In this case, the pebbles pass through the outer region of the core first and each consecutive pass is located in a fuel zone closer to the inner reflector. Thereby, the resulting radial distribution of the fissile material in the core is influenced and the temperature profile is close to optimal.The fuel temperature in the pebbles can be further reduced by reducing the standard pebble diameter from 6 cm to a value as low as 1 cm. An analytical investigation is used to demonstrate the effects on the fuel temperature and pressure drop for both radial and axial cooling.Finally, two-dimensional numerical calculations were performed, using codes for neutronics, thermal-hydraulics and fuel depletion analysis, in order to validate the results for the optimized design that were obtained from the analytical investigations. It was found that for a radially cooled design with an optimized power profile and reduced pebble diameter (below 3.5 cm) both a reduction in the pressure drop ( bar), which increases the reactor efficiency with several percent, and a reduction in the maximum fuel temperature (C) can be achieved compared to present axially cooled designs.     

Gas phase in experimental fuel elements with compact uranium dioxide,irradiated in the SM-2 reactor

Conclusions A complex procedure has been developed for the study of gas release from nuclear fuel, including reactor measurements and post-reactor determination of the amount and composition of the gas medium in the fuel elements at room and elevated temperatures. In fuel elements with compact uranium dioxide (density 10.0–10.43 g/cm³), in addition to gaseous fission products and the helium introduced, Ar, H₂, O₂, CO, CO₂, and N₂ are present, and after irradiation their quantity exceeds the initial quantity, measured for unirradiated fuel elements, by a factor of several.The yield of Xe and Kr under the can of the fuel elements during irradiation of uranium dioxide in the SM-2 reactor amounts to 30–50%, but the measured ratio of Xe/Kr exceeds the calculated ratio by a factor of 1.2, because of the reaction¹³⁵Xe(n, )¹³⁶Xe. The content in the fuel of adsorbed helium is equal to 0.004 n.cm³/g UO₂. The data obtained can be used for physics and technological calculations, and also for refining the procedure for the determination of gas release.Translated from Atomnaya Énergiya, Vol. 57, No. 2, pp. 91–95, August, 1984.     

Determination of the burnout of a fuel element from the icebreaker lenin on the basis of Cs137 activity without chemical separation

It is proposed to use a scintillation gamma spectrometer for determining the absolute values of fuel element burnout. The burnout is determined from the intensity of the gamma lines of the C_s ¹³⁷ (E = 0.66 MeV) accumulated in the fission products.The report gives the results of investigations of U²³⁵ burnout as a function of distance along a depleted fuel element of the reactor on the icebreaker Lenin.The research was conducted in the hot laboratory of the I. V. Kurchatov Atomic Energy Institute.Translated from Atomnaya Énergiya, Vol. 21, No. 2, pp. 92–96, August, 1966.     

Modeling of WWER-440 fuel pin behavior at extended burn-up

Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWERs as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased from 60 to 70 MWd/kg U. The change in the fuel radial power distribution as a function of fuel burn-up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. Both of these features, commonly termed the “rim effect.” High burn-up phenomena in WWER-440 UO₂ fuel pin, which are important for fission gas release (FGR) were modeled. The radial burn-up as a function of the pellet radius and enrichment has to be known to determine the local thermal conductivity.In this paper, the radial burn-up and fissile products distributions of WWER-440 UO₂ fuel pin were evaluated using MCNP4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted FGR calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. A computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented.     

Analysis of radiant heat transfer in a BWR fuel assembly

A computer code ‘CIDER’ was developed which analyzes radiant heat transfer in a BWR fuel rod bundle under loss of coolant conditions. In the code, (1) a channel box and fuel rods are considered to be gray bodies, (2) reflection and absorption of radiation beams in the atmosphere is neglected, (3) a fuel rod is approximated by a regular polygonal rod, and (4) radiant heat flux is calculated considering circumferential temperature distribution on each fuel rod surface, which is determined from radial and circumferential heat conduction calculations in a fuel rod. It was found that the conventional model with uniform cladding temperature overestimated heat flux about 30% in a typical situation, or correspondingly underestimated the temperature rises.     

Irradiation swelling, creep and thermal-stress analysis of LWR fuel elements, computer code isune-2

The irradiation swelling, creep, and thermal-stress analysis of light-water reactor (LWR) oxide (UO₂) fuel elements is analysed. The analysis is based on the basic physical and mathematical assumptions and the experimental data of the fuel and cladding (or canning) materials. In the analysis, the nuclear, physical, metallurgical, and thermo-mechanical properties of the fuel and cladding materials under irradiation environment are examined carefully. The objectives of the paper are mainly (1) to formulate and carry out the irradiation swelling, irradiation creep, and thermal-stress analysis of fuel elements for LWR power reactors, and (2) to develop a computer code which will facilitate the computations for fuel element design, safety analysis, and economic optimization of the power reactors. In a general procedure of the analysis, the irradiation swelling, irradiation creep, temperature distribution, etc. in the fuel and cladding of the oxide fuel elements during the reactor in operation are studied. Some theoretical models and empirical relations (on the basis of accepted experimental data) for irradiation swelling and creep in the fuel and irradiation creep in cladding materials are postulated and developed. Some analytical and empirical relations (based on test results) for heat generation and temperature distribution in the fuel during fuel restructuring are derived. The fuel restructure is, in general, divided into the central void, columnar grain, equiaxed grain, and unaffected grain zones (or regions) after a sufficiently long period for the fuel elements to be irradiated (or operated). From these relations derived for irradiation swelling, irradiation creep, and temperature distribution in the fuel and cladding, together with the well-known strain-stress, incompressibility, compatibility, and stress equilibrium equations, the irradiation swelling, creep, and thermal-stress analysis for the LWR fuel elements can be carried out.From the analytical results obtained, a computer code, ISUNE-2 (which is in the sequence of computer code ISUNE-1 and -1A developed and used previously for liquid-metal fast breeder reactor fuel element design and safety and economic analysis), can be developed. With some reliable experimental data (measured during fuel elements in operation) as input, the computer code may predict various cases of LWR (oxide or carbide) fuel elements in operation. The general scope and resulting contribution of this paper is to provide a realistic analysis and a reliable operating LWR fuel element code for use by nuclear power utilities to predict the fuel element behavior in power reactors. The fuel element design, safety analysis, and economic optimization depend largely on the fuel element behavior in the power reactors.     

Fission-suppressed blankets for fissile fuel breeding fusion reactors

Two blanket concepts for deuterium-tritium (DT) fusion reactors are presented which maximize fissile fuel production while at the same time suppress fission reactions. By suppressing fission reactions, the reactor will be less hazardous, and therefore easier to design, develop, and license. A fusion breeder operating a given nuclear power level can produce much more fissile fuel by suppressing fission reactions. The two blankets described use beryllium for neutron multiplication. One blanket uses two separate circulating molten salts: one salt for tritium breeding and the other salt for U-233 breeding. The other uses separate solid forms of lithium and thorium for breeding and helium for cooling.Nuclear power is the sum of fusion (D + T 14 MeV neutron+ 3.5 MeV alpha) power plus additional power from neutron-induced reactions in the blanket.     

Automated recordkeeping of nuclear materials at electric power plants powered with a water-moderated-water-cooled nuclear reactor

Conclusions The automated nuclear-material recordkeeping system based on the NUMIS-2 program is at present being introduced at the Novororonezh electric power plant. Concurrently, the program is being modified in order to allow its use by the computer model being created in the COMECON member-nations. In the long term, it is expected that the recordkeeping program will become a component part of a single automated control systems of individual electric power plants.The NUMIS-2 program was approved for nuclear-material recordkeeping use after it was tested on the first fuel load of the fourth unit of the Novovoronezh electric power plant. The experiment confirmed the feasibility of automating the processing and storing of nuclear-material information and of recording that information and on the standard computer carriers of information. In principle, these same records can be submitted to the national recordkeeping service (or to the International Atomic Energy Agency) for their further immediate processing by means of the computer program employed by the service. In this manner, the recordkeeping system becomes closed, which makes it promising.We note that the propesed nuclear-material recordkeeping system satisfied the basic requirements of recordkeeping by computer laid down by the International Atomic Energy Agency as well as the required order of reporting by the plants under the control of the Agency.Translated from Atomnaya Énergiya, Vol. 45, No. 4, pp. 267–270, October, 1978.