Thermal-fluid characteristics of the transuranics fuel in a deep-burn HTR core

The Deep Burn Project is evaluating the feasibility of the DB-HTR (Deep Burn High Temperature Reactor) to achieve a very high utilization of transuranics (TRU) derived from the recycle of LWR spent fuel. This study intends to evaluate the thermal-fluid and safety characteristics of TRU fuel in a DB-HTR core using GAMMA+ code. The key design characteristics of the DB-HTR core are more fuel rings (five fuel-rings), less central reflectors (three rings) and decay power curves due to the TRU fuel compositions that are different from the UO₂ fuel. This study considered three types of TRU kernel compositions such as 100%(PuO₂ + NpO₂ + Am), 99.8%(PuO_1.8, NpO₂) + 0.2%UO₂ + 0.6 mole SiC getter, and 70%(PuO_1.8, NpO₂) + 30%UO₂ + 0.6 mole SiC getter. The first fuel type of TRU kernel produces higher decay power than the UO₂ kernel. For the second and the third fuel types, removing the initial Am isotopes and reducing the volumetric packing fraction of TRISO particles will reduce the decay power. The flow distribution, core temperature and TRISO temperature profiles at the steady state were examined. As a safety performance, this study mainly evaluated the peak fuel temperature during LPCC (low pressure conduction cooling) event with considering the impact of decay power, the annealing effect of the irradiated thermal conductivity of graphite, and the impact of the FB (fuel block) end-flux-peaking. For the 600 MW_th DB-HTR core, the peak fuel temperature of 100%(PuO₂ + NpO₂ + Am) TRU was found to be much higher than the transient fuel design limit of 1600 °C due to the lack of heat absorber volume in the central reflector as well as to the increased decay power of the TRU fuel compositions. For a 0.2%UO₂ mixed or a 30%UO₂ mixed TRU, the peak fuel temperature was decreased due to the reduced decay power, however, it was still higher than 1600 °C due to the lack of heat absorber volume in the central reflector.     

Validation of 134Cs, 137Cs and 154Eu single ratios as burnup monitors for ultra-high burnup UO2 fuel

A measurement station has been built for the non-destructive investigation of burnt fuel rod segments through high-resolution gamma spectrometry. Four UO₂ pressurised water reactor fuel rod segments with different burnup levels between 50 and >100 GWd/t and ⩽10 year cooling time have been experimentally characterised using gamma-ray spectrometry to determine ¹³⁴Cs, ¹³⁷Cs and ¹⁵⁴Eu and their corresponding concentration ratios. Experimental errors of ∼2% (1σ) for the ¹³⁴Cs/¹³⁷Cs ratio were obtained for most of the segments. In parallel, pin cell depletion calculations have been performed for each segment using the deterministic code CASMO-4. Measured and calculated ratios have then been compared with the purpose of deriving and validating pin-averaged single-ratio burnup indicators for very high burnups. It is shown that the ¹³⁴Cs/¹³⁷Cs ratio, frequently used as a burnup monitor, is considerably less precise for values exceeding 50 GWd/t; discrepancies of ∼16% are found between measured and calculated values, increasing with burnup up to ∼23%. The ratios built with the ¹⁵⁴Eu concentration show even much larger discrepancies, essentially because this isotope is rather poorly predicted as revealed by just using different basic cross section data.     

The use of NRA to study thermal diffusion of helium in (U, Pu)O2

A lot of work has been already done on helium atomic diffusion in UO₂ samples, but information is still lacking about the fate of helium in high level damaged UOX and MOX matrices and more precisely their intrinsic evolutions under alpha self irradiation in disposal/storage conditions.The present study deals with helium atomic diffusion in actinide doped samples versus damage level. The presently used samples allow a disposal simulation of about 100 years of a UOX spent fuel with a 60 MW d kg^?1 burnup or a storage simulation of a MOX spent fuel with a 47.5 MW d kg^?1 burnup.For the first time, nuclear reaction analysis of radioactive samples has been performed in order to obtain diffusion coefficients of helium in (U, Pu)O₂. Samples were implanted with ³He⁺ and then annealed at temperatures ranging from 1123 K to 1273 K. The evolution of the ³He depth profiles was studied by the mean of the non-resonant reaction: ³He(d, p)⁴He. Using the SIMNRA software and the second Fick’s law, thermal diffusion coefficients have been measured and compared to the ³He thermal diffusion coefficients in UO₂ found in the literature.     

Long-Cycle and High-Burnup Fuel Assembly for the VHTR

For a prismatic VHTR fuel assembly, a physics study has been performed to maximize the fuel performance in terms of the cycle length and the discharge burnup for a given fuel enrichment. The relationship between the fuel performance and the fuel configurations has been investigated in terms of the TRISO packing fraction, diameter of the fuel kernel, fuel management, and moderating power of the fuel block. Both a typical low-enrichment uranium fuel (LEU) and a fuel made of transuranics (TRU) from LWR spent fuel are considered in this paper. It is shown that in order to obtain a long refueling cycle and a high burnup at the same time, the fuel loading needs to be increased together with the moderating power of the fuel block. Three ways are considered for a higher moderation of the fuel block: a larger pitch of the coolant hole pattern, an extra graphite thickness in the fuel block, and a higher graphite density. The impact of the increased pitch on the fuel temperature is also evaluated with a thermal analysis code. We have shown that long refueling cycles and high burnups can be achieved simultaneously for both LEU and TRU fuels.     

Comparative study of actinide and fission product inventory of HEU and potential LEU fuels for MNSRs

A comparative study of fuel burnup and buildup of actinides and fission products for potential LEU fuels (UO₂ and U–9Mo) with existing HEU fuel (UAl₄–Al, 90% enriched) for a typical Miniature Neutron Source Reactor (MNSR) has been carried-out using the WIMSD4 computer program. For the complete burnup, the UAl₄–Al, UO₂ and U–9Mo based systems show a total consumption of 6.89, 6.83 and 6.88 g of ²³⁵U, respectively. Relative to 0.042 g ²³⁹Pu produced in case of UAl₄–Al HEU core, UO₂ and U–9Mo based cores have been found to yield 0.793 and 0.799 g, respectively, indicating much larger values of conversion ratios and correspondingly high values of fuel utilization factor. The end-of-cycle activity of the HEU core has been found 2284 Ci which agrees well with value found by Khattab where as for UO₂ based and U–9Mo based LEU cores show 1.8 and 4.8% increase with values 2326 and 2394 Ci, respectively.     

Determination of high burn-up nuclear fuel elastic properties with acoustic microscopy

We report the measurement of elastic constants of non-irradiated UO₂, SIMFUEL (simulated spent fuel: UO₂ with several additives which aim to simulate the effect of burnup) and irradiated fuel by focused acoustic microscopy. To qualify the technique a parametric study was conducted by performing measurements on depleted uranium oxide (with various volume fraction of porosity, Oxygen-to-metal ratios, grain sizes) and SIMFUEL and by comparing them with previous works presented in the literature. Our approach was in line with existing literature for each parameter studied. It was shown that the main parameters influencing the elastic moduli are the amount of fission products in solution (related to burnup) and the pore density and shape, the influence of which has been evaluated. The other parameters (irradiation defects, oxygen-to-metal ratio and grain sizes) mainly increase the attenuation of the ultrasonic wave but do not change the wave velocity, which is used in the proposed method to evaluate Young’s modulus. Measurements on irradiated fuel (HBRP and N118) were then performed. A global decrease of 25% of the elastic modulus between 0 and 100 GWd/tM was observed. This observation is compared to results obtained with measurements conducted at ITU by Knoop indentation techniques.     

Power ramped cladding stresses and strains in 3D simulations with burnup-dependent pellet–clad friction

This paper presents 2D(r, θ) plane strain and 3D simulations of PCI during base irradiation and ramp tests. Inverse analysis is used to estimate the evolution of friction at the pellet–clad interface with burnup. The number of radial cracks that form during power ramp tests in seventeen UO₂-Zy4 rodlets with burnups in the range 20–60 GWd/tU is the main parameter on which inverse analysis is based. It is shown that the sole evolution of the friction coefficient with burnup is sufficient to capture the radial crack pattern of the rodlets after power ramping. A simple relation between the friction coefficient and the burnup variation after initial pellet–clad contact is thus proposed and used in 3D simulations of PCI. The delayed gap closing at mid-pellet level with respect to inter-pellet level leads to an axial variation of the friction coefficient, with maximum values near the pellet ends. The consequences in terms of PCI failure propensity are then discussed.     

Preparation,characterisation and dissolution of a CeO2 analogue for UO2 nuclear fuel

The behaviour of spent nuclear fuel under geological conditions is a major issue underpinning the safety case for final disposal. This work describes the preparation and characterisation of a non-radioactive UO₂ fuel analogue, CeO₂, to be used to investigate nuclear fuel dissolution under realistic repository conditions as part of a developing EU research programme. The densification behaviour of several cerium dioxide powders, derived from cerium oxalate, were investigated to aid the selection of a suitable powder for fabrication of fuel analogues for powder dissolution tests. CeO₂ powders prepared by calcination of cerium oxalate at 800 °C and sintering at 1700 °C gave samples with similar microstructure to UO₂ fuel and SIMFUEL. The suitability of the optimised synthesis route for dissolution was tested in a dissolution experiment conducted at 90 °C in 0.01 M HNO₃.     

An Apparatus for Continuously Measuring Radon-222 Exhalation from Ground

High burnup MOX and UO₂ test rods were prepared from the fuel rods irradiated in commercial BWRs. Each test rod was equipped with a fuel center thermocouple and reirradiated in the Halden boiling water reactor (HBWR) in Norway. The burnups of MOX and UO₂ test rods reached about 84GWd/tHM and 72GWd/t, respectively. Fuel temperature was measured continuously during the re-irradiation tests. Thermal conductivity change in high burnup fuel was evaluated from the results of comparison between the measured fuel temperature and the data calculated by using the fuel analysis code FEMAXI-6. The comparison results suggested that the thermal conductivity of MOX fuel pellets is comparable to that of UO₂ fuel pellets in the high burnup region around 80 GWd/t. It is probable that the impurity effect of Pu atoms gradually diminishes with increasing burnup because other factors that affect pellet thermal conductivity, such as the accumulation effect of soluble fission products and irradiation-induced defects in crystal lattice, become dominant in a high burnup region.     

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.     

Long life small CANDLE-HTGRs with thorium

CANDLE (constant axial shape of neutron flux, nuclide densities and power shape during life of energy producing reactor) burnup strategy is applied to small (30 MWth) block-type high temperature gas-cooled reactors (HTGRs) with thorium fuel. The CANDLE burnup is adopted in this study since it has several promising merits such as simple and safe reactor operation, and the ease of designing a long life reactor core. Burnup performances of thorium fuel (²³³U, ²³²Th)O₂ are investigated for a range of enrichment ⩽15%. Discharged fuel burnup and burning region motion velocity are major parameters of its performances in this study. The reactors with thorium fuel show a better burnup performance in terms of higher discharged fuel burnup and slower burning region motion velocity (longer core lifetime) compared to the reactors with uranium fuel.     

Characterization of a sodium-cooled fast reactor in an MHR–SFR synergy for TRU transmutation

In the task of destroying the light water reactor (LWR) transuranics (TRUs), we consider the concept of a synergistic combination of a deep-burn (DB) gas-cooled reactor followed by a sodium-cooled fast reactor (SFR), as an alternative way to the direct feeding of the LWR TRUs to the SFR. In the synergy concept, TRUs from LWR are first deeply incinerated in a graphite-moderated DB-MHR (modular helium reactor) and then the spent fuels of DB-MHR are recycled into the closed-cycle SFR. The DB-MHR core is 100% TRU-loaded and a deep-burning (50–65%) is achieved in a safe manner (as discussed in our previous work). In this analysis, the SFR fuel cycle is closed with a pyro-processing technology to minimize the waste stream to a final repository. Neutronic characteristics of the SFR core in the MHR–SFR synergy have been evaluated from the core physics point of view. Also, we have compared core characteristics of the synergy SFR with those of a stand-alone SFR transuranic burner. For a consistent comparison, the two SFRs are designed to have the same TRU consumption rate of ∼250 kg/GW EFPY that corresponds to the TRU discharge rate from three 600 MW DB-MHRs. The results of our work show that the synergy SFR, fed with TRUs from DB-MHR, has a much smaller burnup reactivity swing, a slightly greater delayed neutron fraction (both positive features) but also a higher sodium void worth and a less negative Doppler coefficients than the conventional SFR, fed with TRUs directly from the LWRs. In addition, several design measures have been considered to reduce the sodium void worth in the synergy SFR core.     

Assessment of fuel conversion from HEU to LEU in the Syrian MNSR reactor using the MCNP code

Assessment of fuel conversion from high enriched uranium (HEU) to low enriched uranium (LEU) fuel in the Syrian MNSR reactor was conducted in this paper. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed to assess the possibility of fuel conversion from 89.87% HEU fuel (UAl₄–Al) to 19.75% LEU fuel (UO₂). The first model showed that 347 fuel rods with HEU fuel were required to obtain a reactor core with 5.17 mk unadjusted excess reactivity. The second model showed that only 200 LEU fuel rods distributed in the reactor core like the David star figure were required to obtain a reactor core with 4.85 mk unadjusted excess reactivity. The control rod worth using the LEU fuel was enhanced. Finally, the third model showed that distribution of 200 LEU fuel rods isotropically in the 10 circles of the reactor core failed to convert the fuel since the calculated core unadjusted excess reactivity for this model was 10.45 mk. This value was far beyond the reactor operation limits and highly exceeded the current MNSR core unadjusted excess reactivity (5.17 mk).     

Low cost, proliferation resistant, uranium–thorium dioxide fuels for light water reactors

Our objective is to develop a fuel for the existing light water reactors (LWRs) that, (a) is less expensive to fabricate than the current uranium-dioxide (UO₂) fuel; (b) allows longer refueling cycles and higher sustainable plant capacity factors; (c) is very resistant to nuclear weapon-material proliferation; (d) results in a more stable and insoluble waste form; and (e) generates less high level waste. This paper presents the results of our initial investigation of a LWR fuel consisting of mixed thorium dioxide and uranium dioxide (ThO₂–UO₂). Our calculations using the SCALE 4.4 and MOCUP code systems indicate that the mixed ThO₂–UO₂ fuel, with about 6 wt.% of the total heavy metal U-235, could be burned to 72 MW day kg⁻¹ (megawatt thermal days per kilogram) using 30 wt.% UO₂ and the balance ThO₂. The ThO₂–UO₂ cores can also be burned to about 87 MW day kg⁻¹ using 35 wt.% UO₂ and 65% ThO₂with an initial enrichment of about 7 wt.% of the total heavy metal fissile material. Economic analyses indicate that the ThO₂–UO₂ fuel will require less separative work and less total heavy metal (thorium and uranium) feedstock. At reasonable future costs for raw materials and separative work, the cost of the ThO₂–UO₂ fuel is about 9% less than uranium fuel burned to 72 MW day kg⁻¹. Because ThO₂–UO₂ fuel will operate somewhat cooler, and retain within the fuel more of the fission products, especially the gasses, ThO₂–UO₂ fuel can probably be operated successfully to higher burnups than UO₂ fuel. This will allow for longer refueling cycles and better plant capacity factors. The uranium in our calculations remained below 20 wt.% total fissile fraction throughout the cycle, making it unusable for weapons. Total plutonium production per MW day was a factor of 3.2 less in the ThO₂–UO₂ fuel than in the conventional UO₂ fuel burned to 45 MW day kg⁻¹. Pu-239 production per MW day was a factor of about 4 less in the ThO₂–UO₂ fuel than in the conventional fuel. The plutonium produced was high in Pu-238, leading to a decay heat about three times greater than that from plutonium derived from conventional fuel burned to 45 MW day kg⁻¹ and 20 times greater than weapons grade plutonium. This will make fabrication of a weapon more difficult. Spontaneous neutron production from the plutonium in the ThO₂–UO₂ fuel was about 50% greater than that from conventional fuel and ten times greater than that from weapons grade plutonium. High spontaneous neutron production drastically limits the probable yield of a crude weapon. Because ThO₂ is the highest oxide of thorium while UO₂ can be oxidized further to U₃O₈ or UO₃, ThO₂–UO₂ fuel appears to be a superior waste form if the spent fuel is to be exposed ever to air or oxygenated water. And, finally, use of higher burnup fuel will result in proportionally fewer spent fuel bundles to handle, store, ship, and permanently dispose of.     

Fuel performance analysis for PWR cores

Attainable discharge burnups for oxide and hydride fuels in PWR cores were investigated using the TRANSURANUS fuel performance code. Allowable average linear heat rates and coolant mass fluxes for a set of fuel designs with different fuel rod diameters and pitch-to-diameter ratios were obtained by VIPRE and adopted in the fuel code as boundary conditions. TRANSURANUS yielded the maximum rod discharge burnups of the several design combinations, under the condition that specific thermal-mechanical fuel rod constraints were not violated. The study shows that independent of the fuel form (oxide or hydride) rods with (a) small diameters and moderate P/Ds or (b) large diameters and small P/Ds give the highest permissible burnups limited by the rod thermal-mechanical constraints. TRANSURANUS predicts that burnups of ∼74 MWd/kg U and ∼163 MWd/kg U (or ∼65.2 MWd/kg U oxide-equivalent) could be achieved for UO₂ and UZrH_x fuels, respectively. Furthermore, for each fuel type, changing the enrichment has only a negligible effect on the permissible burnup. The oxide rod performance is limited by internal pressure due to fission gas release, while the hydride fuel can be limited by excessive clad deformation in tension due to fuel swelling, unless the fuel rods will be designed to have a wider liquid metal filled gap. The analysis also indicates that designs featuring a relatively large number of fuel rods of relatively small diameters can achieve maximum burnup and provide maximum core power density because they allow the fuel rods to operate at moderate to low linear heat rates.     

快堆中不同核燃料类型的长寿期性能评价

建立了典型的快堆六角形栅元堆芯模型,研究了多种类型的燃料在快中子能谱辐照环境下经过较长时间辐照后的性能,对不同燃料堆芯在运行寿期末的乏燃料组成成分进行了分析.结果表明,在栅元结构完全一样且初始剩余反应性基本相同的情况下,燃料反应性损失从小到大的顺序是:金属燃料＜氮化物燃料＜碳化物燃料<氧化物燃料;在整个寿期中,使用Pu驱动的燃料比使用235U驱动的燃料反应性下降得慢;金属燃料寿期末乏燃料中按初始装载燃料质量平均后的超铀核素的质量最小,其他依次为氧化物<氮化物<碳化物;由于初始装载量的增多,使用Pu驱动的燃料寿期末乏燃料超铀核素的总量比使用235U驱动的燃料多,同时,乏燃料Pu中的易裂变同位素的份额比235U驱动燃料的少.从中子学角度考虑,UZr燃料是比较理想的长寿命快堆堆芯燃料类型.     

Some aspects of risk reduction strategy by multiple recycling in fast burner reactors of the plutonium and minor actinide inventories

This paper shows the impact of recycling light water reactor (LWR) mixed oxide (MOX) fuel in a fast burner reactor on the plutonium (Pu) and minor actinide (MA) inventories and on the related radioactivities. Reprocessing of the targets for multiple recycling will become increasingly difficult as the burnup increases. Multiple recycling of Pu + MA in fast reactors is a feasible option which has to be studied very carefully: the Pu (except the isotopes Pu-238 and Pu-240), Am and Np levels decrease as a function of the recycle number, while the Cm-244 level accumulates and gradually transforms into Cm-245. Long cooling times (10 + 2 years) are necessary with aqueous processing. The paper discusses the problems associated with multiple reprocessing of highly active fuel types and particularly the impact of Pu-238, Am-241 and Cm-244 on the fuel cycle operations. The calculations were performed with the zero-dimensional ORIGEN-2 code. The validity of the results depends on that of the code and its cross-section library. The time span to reduce the initial inventory of Pu + MA by a factor of 10 amounts to 255 years when average burnups are limited to 150 GW · d t⁻¹ (tonne).     

Reduction of fission gas release at extended burnups in light water reactors by decreasing fuel temperature

The effects of fuel temperature on fission gas release in light water reactor UO₂ fuel at extended burnups of up to 56 effective full power months (EFPMs) are evaluated using a simple fission gas release mechanistic model. The model is first described and then model validation comparisons are made against experimental fission gas release date. The study shows that by decreasing the maximum operating fuel temperature to below 1200°C, it is possible to reduce the amount of released fission gas at 56 EFPMs to less than that at the current design burnup of 36 EFPMs.     

Lowering the enrichment of the Syrian miniature neutron source reactor

An investigation of lowering the fuel enrichment of MNSR was realized. A 3-D neutronic model was developed for the analysis of the reactor. It was found that lower number of fuel elements is needed when UO₂ is used with 5.45 g of ²³⁵U content in each fuel rod. Sensitivity of the reactor to the purity of the beryllium reflector proved to be an important factor in determining the reactor neutronics as well as the weight of loaded fuel in the core. Inherent safety features of low excess reactivity and shutdown margins are preserved and enhanced. Average thermal fluxes over different zones of the core are kept very much unchanged. ©     

田湾核电站乏燃料水池采用燃耗信任制的计算研究

以田湾核电站(TNPS)2×5排列的贮存格架构成的乏燃料水池为例,研究采用燃耗信任制技术的密集贮存和临界安全问题。采用MONK9A程序计算分析不同富集度、不同燃耗的乏燃料装载情况下系统的k_eff. 根据系统k_eff随不同初始富集度燃料的燃耗变化情况给出了水池的参考装载曲线。采用燃耗信任制技术的密集贮存方案能提高贮存能力31%。