HTGR fuel and fuel cycle

This paper describes some of the basic charactritics of the HTGR fuel with emphasis on the 1160 MW(e) plant now being offered commercially by Gulf General Atomic and some of the aspects of the fuel cycle which are unique to the HTGR. The fuel cycle is based on highly enriched (93%) uranium for the initial and the make-up fissile material; thorium for the fertile material, with the bred ²³³U being recycled at the earliest opportunity. The fuel elements consist only of ceramic materials with the thorium/uranium carbides or oxides in the form of coated particles.     

Plutonium fuel and fuel elements for power reactors

Conclusions The use of plutonium in the fuel cycle during complex utilization of thermal and fast reactors in nuclear energetics permits solving the problem of ensuring nuclear fuel for a long period. Oxide uranium-plutonium fuel facilitates the development of technology of fast reactors and so far it is considered as the basic type of fuel. At the same time, oxide fuel cannot ensure the required rate of plutonium accumulation, in view of which the investigations of more efficient fuel and constructional materials become a pressing problem. The use of uranium-plutonium oxide fuel in thermal reactors requires improvements in the construction of fuel elements and organization of large-scale completely automatic production.Translated from Atomnaya Énergiya, Vol. 43, No. 5, pp. 412–417, November, 1977. Editors' Remarks. For the completeness of the discussion of the problem it is, of course, necessary to consider the possibility of using plutonium in fast and thermal reactors as done by the authors. However, it should be kept in mind that by its nuclear-physical parameters plutonium as a nuclear fuel is more suitable for use in fast reactors than in thermal reactors. The use of plutonium in thermal reactors can reduce the demands of natural uranium for the development of nuclear power in all by 10–15%, whereas its use in fast reactors reduces the demand for uranium by a factor of 10.All this indicates the feasibility of using plutonium only in fast reactors even if its accumulation is required over a certain period.     

Investigating fuel element melt in simulators with fuel compositions

Nizhegorod Polytechnical Institute. Translated from Atomnaya Énergiya, Vol. 70, No. 2, pp. 108–110, February, 1991.     

Complex radiation monitoring of the fuel distribution in vibration-packed fuel elements

Conclusions The monitoring procedure, based on the combined use of a number of radiation methods, gives the possibility of obtaining complete information about the distribution of fuel materials in a fuel element. It is expedient to use the -absorption method in conjuction with passive gamma-scanning and computer tomography.Parametric modeling of the monitoring process on a computer allows the characteristics of the monitoring apparatus to be optimized, and its errors to be analyzed.The MNMG-1M apparatus can be used for monitoring the distribution of vibration-packed fuel in fuel-element rods at both the stage of development and research and also in the conditions of their commerical production.Translated from Atomnaya Énergiya, Vol. 59, No. 1, pp. 22–27, July, 1985.     

Effect of fuel — Cladding axial asymmetry on nuclear fuel element operation

Institute of Nuclear Reactors, Russian Scientific Center“Kurchatovskii institut.”Translated from Atomnaya énergiya, Vol. 76, No. 5, pp. 417–422, May, 1994.     

Growth of the interaction layer around fuel particles in dispersion fuel

Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl_x. The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel.     

Recycling of uranium from spent RBMK fuel in the nuclear fuel cycle

It is shown that there is promise in using the uranium product obtained by reprocessing spent nuclear fuel from RBMK reactors as a non-initial fuel source for thermal reactors. A technical path for spent nuclear fuel from RBMK reactors is proposed: radiochemical reprocessing and obtaining oxides of recycled uranium. Oxides of the category RBMK-poor are packed and then stored in a near-surface storage facility; oxides of the category RBMK-rich are fluoridated, and UF₆ is fed into separation production for additional enrichment to the required content of ²³⁵U. Additional advantages of recycled RBMK uranium as a source of non-initial ²³⁵U are the low content of ²³²U and the relatively low activity of spent fuel, which simplifies its reprocessing.     

Radiostability of fuel plates used in SM-2 reactor fuel elements

Conclusions The data cited imply a high structural and dimensional stability on the part of plate type fuel elements used in the SM-2 reactor. The linearity of the volume increase with burnup allows us to assess changes in the physical, hydraulic, and heat-transfer parameters of the core. We note that a process of layer separation resulting in a loss of pressuretightness sets-in in some fuel elements with 30% uranium burnup, and this could be involved in failures during cooling or cooldown. Checking out this last point would enable us to settle on a criterion for the limiting allowable burnup. It is highly probable that this criterion will be related to the physics of the reactor, rather than to the degree of radiation damage of the reactor materials.Translated from Atomnaya Énergiya, Vol. 24, No. 5, pp. 432–435, May, 1968.     

Nitration by a simulated fuel technique for nitride fuel re-fabrication

Nitration reaction of a spent nuclear oxide fuel through a carbothermic reduction and the change in thermal conductivity of the resultant nitride specimens were investigated by a simulated fuel technique for use in nitride fuel re-fabrication from spent oxide fuel. The simulated spent oxide fuel was formed by compacting and sintering a powder mixture of UO₂ and stable oxide fission product impurities. It was pulverized by a 3-cycle successive oxidation-reduction treatment and converted into nitride pellet specimens through the carbothermic reduction. The rate of the nitration reaction of the simulated spent oxide fuel was decreased due to the fission product impurities when compared with pure uranium dioxide. The amount of Ba and Sr in the simulated spent oxide fuel was considerably reduced after the nitride fuel re-fabrication. The thermal conductivity of the nitride pellet specimen in the range 295-373 K was lower than that of the pure uranium nitride but higher than the simulated spent oxide fuel containing fission product impurities.     

Effect of oxide layer growth on fuel temperature of mini fuel plates

The effect of the oxide layer formed on the mini fuel plates is studied to evaluate fuel centerline temperature. For a part of U-Mo fuel qualification program, mini fuel plates, double-stacked as upper and lower plates, will be irradiated in the HANARO reactor for four cycles. In the present study, fuel performance and thermal hydraulic behavior during irradiation are numerically investigated using the MCNP and TMAP codes. The power released from the mini fuel plates is estimated using the MCNP code. From the neutronic analysis results, it is observed that the lower plate at the BOC during the 1st cycle releases the highest power, and the power gradually decreases during the irradiation test. The growth of the oxide layer thickness during the irradiation test is predicted using many correlations with various pH values ranging from 5.0 to 7.0. The pH value in the HANARO reactor is controlled between 5.7 and 6.2, and the oxide layer thickness is predicted by the Boehmite model for these two pH values. The oxide layer thickness predicted using the other correlations are bounded by these two predicted values. The maximum oxide layer thickness at the end of irradiation is approx. 9 and 68 μm with pH of 5.7 and 6.2, respectively. The Pawel model with a rate factor of 16 predicts the maximum oxide layer as 25 μm. Using the predictions of the oxide layer thickness, the centerline fuel temperatures are evaluated using the TMAP code. The maximum fuel temperature is not observed when the power released from the fuel is the highest. Because the temperature rise through the oxide layer is significant, the oxide layer thickness must be considered in the fuel temperature evaluation. The oxide formation saturates with time, and the fuel reaches the maximum temperature at the end of the saturation. After the maximum fuel temperature is reached, it starts decreasing, because the power decreases.     

Thermal deformation of fuel elements

The authors calculate the temperature drops in the walls of cassette-type fuel elements under neutron fluxes with radial gradients. The thermal deformations of the wails are measured in the working range of temperature drops.Translated from Atomnaya Énergiya, Vol. 21, No. 1, pp. 22–26, July, 1966.     

Subchannel analysis of CANDU-SCWR fuel

A cooperative study has been initiated at Xi'an Jiaotong University (XJTU) with Atomic Energy of Canada Limited (AECL) to develop a subchannel code ATHAS for preliminary analyses of flow and enthalpy distributions and cladding temperatures in CANDU fuel at super-critical water conditions. The code is applicable for transient and steady-state calculations. Then the paper uses the ATHAS code to analyze CANDU-SCWR which is operating at 25.0 MPa pressure. The results show that the maximum cladding-surface temperature of CANFLEX bundle is 804.1 °C, which is below the limit of design, and it is appropriate for use in the CANDU super-critical water-cooled reactor (SCWR) based on heat-transfer analysis.     

Uranium-zirconium hydride fuel properties

Properties of the two-phase hydride U_0.3ZrH_1.6 pertinent to performance as a nuclear fuel for LWRs are reviewed. Much of the available data come from the Space Nuclear Auxiliary Power (SNAP) program of 4 decades ago and from the more restricted data base prepared for the TRIGA research reactors some 3 decades back. Transport, mechanical, thermal and chemical properties are summarized. A principal difference between oxide and hydride fuels is the high thermal conductivity of the latter. This feature greatly decreases the temperature drop over the fuel during operation, thereby reducing the release of fission gases to the fraction due only to recoil. However, very unusual early swelling due to void formation around the uranium particles has been observed in hydride fuels. Avoidance of this source of swelling limits the maximum fuel temperature to ∼650 °C (the design limit recommended by the fuel developer is 750 °C). To satisfy this temperature limitation, the fuel-cladding gap needs to be bonded with a liquid metal instead of helium. Because the former has a thermal conductivity ∼100 times larger than the latter, there is no restriction on gap thickness as there is in helium-bonded fuel rods. This opens the possibility of initial gap sizes large enough to significantly delay the onset of pellet-cladding mechanical interaction (PCMI). The large fission-product swelling rate of hydride fuel (3× that of oxide fuel) requires an initial radial fuel-cladding gap of ∼300 m if PCMI is to be avoided. The liquid-metal bond permits operation of the fuel at current LWR linear-heat-generation rates without exceeding any design constraint. The behavior of hydrogen in the fuel is the source of phenomena during operation that are absent in oxide fuels. Because of the large heat of transport (thermal diffusivity) of H in ZrH_x, redistribution of hydrogen in the temperature gradient in the fuel pellet changes the initial H/Zr ratio of 1.6 to ∼1.45 at the center and ∼1.70 at the periphery. Because the density of the hydride decreases with increasing H/Zr ratio, the result of H redistribution is to subject the interior of the pellet to a tensile stress while the outside of the pellet is placed in compression. The resulting stress at the pellet periphery is sufficient to overcome the tensile stress due to thermal expansion in the temperature gradient and to prevent radial cracking that is a characteristic of oxide fuel. Several mechanisms for reduction of the H/Zr ratio during irradiation are identified. The first is transfer of impurity oxygen in the fuel from Zr to rare-earth oxide fission products. The second is the formation of metal hydrides by these same fission products. The third is by loss to the plenum as H₂.The review of the fabrication method for the hydride fuel suggests that its production, even on a large scale, may be significantly higher than the cost of oxide fuel fabrication.     

Thermodynamics of high-temperature nuclear fuel

A method for performing a thermodynamic analysis of the high-temperature nuclear fuel using the ASTA computer program is substantiated. Calculations of the chemical composition and pressure of the gas phase of the ternary systems U-O-C and Pu-O-C are performed. The results obtained are compared with existing experimental data and theoretical studies performed by other authors. The results show that the entropy factor apparently plays an appreciable role in the thermodynamics of the systems studied. A comparative analysis of micropellets with uranium and plutonium fuel is performed. An estimate of the diffusion kinetics of the chemical interaction in micropellets is given as substantiation of the application of the methods of equilibrium thermodynamics for calculating the chemical and phase composition of the nuclear fuel.__________Translated from Atomnaya Énergiya, Vol. 98, No. 1, pp. 36–44, January, 2005.     

Controlled porosity reactor fuel

Thermally stable UO₂ reactor fuel pellets have been obtained using volatile additive pore formers and high activity powders. Various additives, including uranyl and ammonium salts, have been studied, nearly all of which appear similar in effect. Due to the inherently small percentage of fine pores (<2 μm dia.) present and large grain size of pellets made by this process, it is expected that the fuel will exhibit excellent irradiation stability. Density can be controlled reproducibly by varying the amount of pore former such that fuels suitable for either LWR's or fast reactors can be prepared. The size fraction of additive does not affect the thermal response of the pellet. The size distribution of the pores can be controlled within fine limits.     

Thermophysical properties of irradiated uranium-zirconium fuel

The dependence of the thermophysical properties of metallic nuclear fuel — the alloy Zr-40U — in a wide temperature range on the amount of fission products accumulated is presented. Non-irradiated and irradiated samples with different degree of accumulation of fission products — 0.4, 0.6, and 0.9 g/cm³ — are investigated. The specific heat is measured in the range 50–1000°C, the temperature diffusivity is measured in the range 300–1000°C, and the variation of the dimensions and density of the samples on heating is also investigated. The thermal conductivity in the range 50–1000°C is calculated on the basis of the experimental data. __________ Translated from Atomnaya énergiya, Vol. 108, No. 1, pp. 6–9, January, 2008.     

Improving RBMK fuel loading

Translated from Atomnaya Énergiya, Vol. 71, No. 1, pp. 3–8, July, 1991     

Influence of pellet length on interaction between fuel and cladding in simulated reactor fuel pins

An electrically heated fuel pin test apparatus has been developed for out-of-pile investigations of fuel pin parameters with a view to supplementing in-pile experiments. Sixty per cent of reactor heat ratings has been achieved with a hollow pin having an axially located electrical heater, the limitation being the melting of the UO₂ pellets. The theoretical unconstrained shapes of a heated pellet and a fuel pellet under elastic conditions were calculated. Both showed an ‘hour glass’ form suggesting that permanent circumferential ridges would occur in the cladding of a heated pin as they do in the cladding of fuel pins. These ridges were subsequently produced in heated pins, the pins being heated while immersed in cooling water at typical reactor temperatures and pressures. From a series of such tests using different pellet lengths it was found that a significant reduction in ridge height occured when the pellet ratio was one-third of the value in a typical reactor. The temperatures reached in the UO₂ pellets were estimated from a metallographic examination of a pin cross section after test. Using published data of ∫kdT for UO₂ over various temperature ranges the pin heat output at that cross section was determined.     

Oxidation of accident tolerant fuel candidates

In this study, the oxidation of various accident tolerant fuel candidates produced under different conditions have been evaluated and compared relative to the reference standard – UO₂. The candidates considered in this study were UN, U₃Si₂, U₃Si₅, and a composite material composed of UN–U₃Si₂. With the spark plasma sintering (SPS) method, it was possible to fabricate samples of UN with varying porosity, as well as a high-density composite of UN–U₃Si₂?(10%). Using thermogravimetry in air, the oxidation behaviors of each material and the various microstructures of UN were assessed. These results reveal that it is possible to fabricate UN to very high densities using the SPS method, such that its resistance to oxidation can be improved compared to U₃Si₅ and UO₂, and compete favorably with the principal ATF candidates, U₃Si₂, which shows a particularly violent reaction under the conditions of this study, and the UN–U₃Si₂?(10%) composite.     

Validation of frey for the safety analysis of LWR fuel using transient fuel rod experiments

The analysis and comparison of severe light water reactor transient experiments are presented from the FREY verification and validation effort. The purpose of this study was to validate the predictive capabilities of the code for severe transient analysis. The FREY code, developed under the sponsorship of the Electric Power Research Institute, uses a two-dimensional finite-element computational method for the thermomechanical analysis of LWR fuel rods under steady state and transient conditions. A total of 10 test fuel rods from experimental programs conducted in both the Power Burst Facility and the Transient Reactor Test Facility have been used in this study. The fuel rods were selected from the following test programs: Power Coolant Mismatch Tests, PCM-2 and PCM-4: Reactivity Initiated Accident Test, RIA 1–2; Loss-of-Coolant Accident Test, LOC-3; First Fuel Rod Failure Test, FRF-1; and Irradiation Effects Test, IE-3. The test programs used in this study cover a large range of code applications for severe transient analysis. The methods used to model the fuel, cladding, and coolant geometry are discussed in addition to experimental data comparisons. The results of the PCM-2, RIA 1–2, and FRF-1 analyses are presented to highlight the full two-dimensional modeling capabilities of FREY and to compare the thermal and mechanical measurements with FREY's prediction. The comparisons show good general agreement, with a tendency for FREY to overpredict the peak cladding surface temperature for a few cases where strong three-dimensional effects have been identified.