The objective of this study was to evaluate accident-tolerant fuel (ATF) concepts being considered for CANDU reactors. Several concepts, including uranium dioxide/silicon carbide (UO2-SiC) composite fuel, dense fuels, microencapsulated fuels, and ATF cladding, were modelled in Serpent 2 to obtain reactor physics parameters, including important feedback parameters such as coolant void reactivity and fuel temperature coefficient. In addition, fuel heat transfer was modelled, and a simple accident model was tested on several ATF cases to compare with UO2. Overall, several concepts would require enrichment of uranium to avoid significant burnup penalties, particularly uranium-molybdenum (U-Mo) and fully ceramic microencapsulated (FCM) fuels. In addition, none of the fuel types have a significant advantage over UO2 in terms of overall accident response or coping time, though U-9Mo fuel melts significantly sooner due to its low melting point. Instead, the different ATF concepts appear to have more modest advantages, such as reduced fission product release upon cladding failure, or reduced hydrogen generation, though a proper risk assessment would be required to determine the magnitude of these advantages to weigh against economic disadvantages. The use of uranium nitride (UN) enriched in
While there are many studies evaluating the performance of accident-tolerant fuels (ATF) in light water reactors (LWRs), there are comparatively few studies which look at the use of ATF in CANDU reactors. ATF has the potential to reduce the consequences of a severe accident and/or give the operators more time to mitigate the consequences of a severe accident. However, as there are significant differences between LWRs and CANDU reactors in terms of reactor and fuel design, ATF should be evaluated specifically for CANDU reactors in order to determine their viability for future use in CANDU reactors.
This study in particular focused primarily on reactor physics, using Serpent 2 to evaluate the CANDU fuel lattice behaviour for different ATF loadings. While thermal hydraulics were not specifically modelled for this study, heat transfer from the fuel is a key consideration for accident-tolerant fuel and has a significant effect on the reactor physics; thus simple heat transfer models were constructed in order to calculate fuel temperatures for each case. The accident-tolerant fuel cases include the following: Uranium dioxide/silicon carbide (UO2-SiC) composite fuel High-density fuels: uranium nitride (UN), uranium silicide (U3Si2) (as an additive to UN), and uranium-molybdenum alloy fuel (U-9Mo) Fully ceramic microencapsulated (FCM) fuels Accident-tolerant cladding materials.
Most power reactors rely on fuel consisting of uranium dioxide pellets surrounded by a zircaloy sheath. UO2 is preferred over uranium metal due to greater stability and a high melting point (roughly 2850°C) but has a disadvantage of a poor thermal conductivity, particularly at higher temperatures and irradiation levels [
Accident progression in CANDU reactors differs from that in LWRs. In the case of a loss-of-coolant accident (LOCA) or station blackout (SBO) where cooling is degraded, heat removal may still occur via radiation to the pressure tube. As the pressure tube heats up, it will balloon or sag, depending on system pressure, until it touches the calandria tube, in which case heat can be conducted to the moderator. The large volume of water in the moderator can then sustain a heat sink for a prolonged period, providing operators with a significant window to implement emergency procedures [
Studies of accident-tolerant fuels have increased since the Fukushima disaster, where a SBO scenario led to fuel cladding oxidation, fuel melting, hydrogen explosions, and the release of radiation into the environment [
The simplest proposal for accident-tolerant fuel is to mix up to 10% silicon carbide by volume into UO2, forming composite pellets through spark plasma sintering [
Uranium nitride has been proposed as an accident-tolerant fuel due to a much greater thermal conductivity and greater uranium density compared to UO2, while still having a high melting point. Typical impurity levels can be kept well below 1% [
Uranium silicide (U3Si2) has been proposed as a compromise between UO2 and UN, with uranium density and thermal conductivity between the two, along with a neutron economy similar to UO2 and less reactivity with water compared to UN, though with a lower melting point than either. Uranium silicide can also be mixed into uranium nitride to combine their properties. The uranium silicide would protect the uranium nitride from exposure, while still providing a better thermal conductivity than UO2 [
Another proposal is for uranium-molybdenum alloys, with a high uranium density, good thermal conductivity, and a fair neutron economy, though with a rather low melting temperature. With a melting point of only around 1400 K [
One final option for the fuel pellets is to manufacture fully ceramic microencapsulated (FCM) fuel. In this fuel, uranium dioxide or uranium nitride is embedded within tristructural-isotropic (TRISO) particles, which use several layers, including a silicon carbide layer, to contain fission products [
The thermal conductivity of SiC is particularly susceptible to irradiation-induced degradation [
In terms of cladding, modifications fall under three categories: metallic cladding, ceramic cladding (SiC), and multilayer claddings or surface coatings. Replacing the zircaloy cladding with another alloy such as stainless steel or iron-chromium-aluminum alloy (FeCrAl) [
The zircaloy cladding can also be replaced by a SiC-based cladding. Unlike metals, SiC is virtually transparent to neutrons and also has a significantly higher melting point than stainless steels. However, silicon carbide is relatively brittle compared to metals, increasing the risk of failure due to thermal and mechanical stresses. Some proposals have been made to mitigate this issue, such as producing SiC fibres rather than monolithic SiC and embedding the fibres to form a composite, though this has the disadvantage of being permeable to fission products [
The final option is simply coating the zircaloy with an oxidation-resistant material, either metallic or ceramic. This provides the oxidation resistance of the surface material while retaining the mechanical properties of the zircaloy and greatly reducing the impact on neutron economy as less neutron-absorbing material is used. The burnup penalty is minimal when compared to monolithic FeCrAl [
The following properties for accident-tolerant fuel materials have been identified in Tables
Uranium density comparison for UO2 and ATF fuel.
Fuel material | Uranium density |
Ref. |
---|---|---|
UO2 (fully dense) | 9.66 | [ |
UO2 (5% porosity) | 9.18 | |
UN (fully dense) | 13.55 | [ |
U3Si2 (fully dense) | 11.31 | [ |
U-8Mo (fully dense) | 16.1 | [ |
FCM UO2/SiC (55% TRISO packing) | 1.84 | [ |
FCM UN/SiC (55% TRISO packing) | 2.72 | [ |
Melting points for UO2 and ATF fuel materials.
Fuel material | Melting point (°C) | Ref. |
---|---|---|
UO2 | 2847 | [ |
UN | 2680 | [ |
U3Si2 | 1665 | [ |
U-8Mo | 1135 | [ |
SiC | 2457 | [ |
Comparison of thermal conductivity of UO2 and ATF materials [
The fuels evaluated in this study were as follows: UO2/SiC composite fuel Uranium nitride fuel (enriched to 99.5% UN/U3Si2 composite fuel UN fuel mixed with zirconium nitride (as U-9Mo fuel FCM with UO2 and UN in SiC, assuming bare kernels embedded in the matrix to maximize fuel loading, thus acting more like a microcell fuel FCM with UN TRISO particles in SiC Silicon carbide cladding Two-layer cladding using zircaloy with a 20% thickness FeCrAl layer. The cladding consists of a zircaloy layer of 80% of the nominal CANDU cladding thickness, with the FeCrAl coating thickness of 20% of the nominal CANDU cladding thickness, so that the total thickness is the same as for the standard zircaloy cladding.
The full case list is shown in Table
ATF case table analyzed.
Case | Fuel | Cladding | Target BU |
---|---|---|---|
(MWh/kgU) | |||
|
Uranium dioxide | Zircaloy-4 | 200 (NU) |
|
Uranium dioxide | Zircaloy-4 | 600 |
|
UO2 + 3% SiC by volume | Zircaloy-4 | NU, 200, 600 |
|
UO2 + 3% SiC by volume | SiC | 200 |
|
UO2 + 3% SiC by volume | Zr/FeCrAl | 200 |
|
UO2 + 6% SiC by volume | Zircaloy-4 | 200 |
|
UO2 + 10% SiC by volume | Zircaloy-4 | 200 |
|
Uranium nitride | Zircaloy-4 | NU, 600 |
|
UN + 3% U3Si2 by volume | Zircaloy-4 | NU |
|
UN + 6% U3Si2 by volume | Zircaloy-4 | NU |
|
UN + 10% U3Si2 by volume | Zircaloy-4 | NU |
|
UN + 10% ZrN by volume | Zircaloy-4 | NU |
|
U-9Mo | Zircaloy-4 | 200 |
|
FCM, UO2 kernels in SiC, 40% packing, 700 |
Zircaloy-4 | NU, 200, 600, |
|
FCM, UO2 kernels in SiC, 45% packing, 700 |
Zircaloy-4 | 200, |
|
FCM, UO2 kernels in SiC, 35% packing, 700 |
Zircaloy-4 | 200, |
|
FCM, UO2 kernels in SiC, 40% packing, 650 |
Zircaloy-4 | 200, |
|
FCM, UO2 kernels in SiC, 40% packing, 750 |
Zircaloy-4 | 200, |
|
FCM, UO2 kernels in SiC, 50% packing, 700 |
Zircaloy-4 | 200, |
|
FCM, UO2 kernels in SiC, 55% packing, 700 |
Zircaloy-4 | 200, |
|
FCM, UN kernels in SiC, 40% packing, 700 |
Zircaloy-4 | 200, |
|
FCM, UN kernels in SiC, 45% packing, 700 |
Zircaloy-4 | 200, |
|
FCM, UN kernels in SiC, 35% packing, 700 |
Zircaloy-4 | 200, |
|
FCM, UN kernels in SiC, 50% packing, 700 |
Zircaloy-4 | 200, |
|
FCM, UN kernels in SiC, 55% packing, 700 |
Zircaloy-4 | 200, |
|
FCM, UN TRISO in SiC, 55% packing, 700 |
Zircaloy-4 |
|
Several different exit burnups were evaluated. Nearly all cases were evaluated for an exit burnup of 200 MWh/kgU. As the calculations were for a lattice and not a full core, an average excess reactivity was calculated for natural UO2 and the exit burnup for all other cases calculated based on this excess reactivity. Some cases were also computed at a higher exit burnup of 600 MWh/kgU. However, for an exit burnup of 200 MWh/kgU, the average residence time of the fuel is proportional to the fuel’s uranium density. This is of particular concern for FCM fuel, which has a very low uranium density. As 200 MWh/kgU is equivalent to 4044 MWh/bundle for UO2 fuel as modelled in this study, cases for FCM fuel with an exit burnup of 4044 MWh/bundle were added.
For this study, the Serpent 2 code [
CANDU lattice properties.
Element | Property | Units | Value | Comment | |||
---|---|---|---|---|---|---|---|
Fuel | Temperature | K | 941.29 | ||||
Diameter | cm | 1.2244 | |||||
|
|||||||
Cladding | Temperature | K | 560.66 | ||||
OD |
cm | 1.308 | |||||
|
|||||||
Fuel bundle | Type | 37-element | |||||
Number of pins | 1 | 6 | 12 | 18 | |||
Ring radii | cm | 0 | 1.5 | 2.9 | 4.35 | ||
Ring pitches | ° | 0 | 0 | 15 | 0 | ||
|
|||||||
Coolant | Composition | D2O | 99.2% purity | ||||
Density | g/cm3 | 0.81212 | |||||
Temperature | K | 560.66 | |||||
|
|||||||
Pressure tube | Composition | Zr-Nb Alloy | 97.5 wt% Zr | ||||
Density | g/cm3 | 6.57 | |||||
Temperature | K | 560.66 | |||||
ID |
cm | 10.3378 | |||||
OD | cm | 11.2064 | |||||
|
|||||||
Gas Gap | Composition | CO2 | |||||
Density | g/cm3 | 0.002297 | |||||
Temperature | K | 345.66 | |||||
|
|||||||
Calandria tube | Composition | Zr Alloy | 97.24 wt% Zr | ||||
Density | g/cm3 | 6.44 | |||||
Temperature | K | 345.66 | |||||
ID | cm | 12.8956 | |||||
OD | cm | 13.1750 | |||||
|
|||||||
Moderator | Composition | D2O | 99.91% purity | ||||
Density | g/cm3 | 1.082885 | |||||
Temperature | K | 345.66 | |||||
|
|||||||
Lattice pitch | cm | 28.575 |
The research steps are shown in Figure
Methodology process.
To calculate the enrichment required for each case, the excess reactivity of each case is defined as follows:
The integral was computed for each burnup case using the trapezoid method. An excess reactivity of roughly 35 mk was calculated for the base UO2 case. Therefore, the enrichments of other cases were varied until they came within 1-2 mk of the same 35 mk excess reactivity, which accounts for leakage and reactivity devices.
The following equation was used to compute the concentration of 234U with respect to enrichment [
The ENDF/B-VII.0 libraries provided with Serpent were used, and Doppler Broadening Rejection Correction (DBRC) [
Effect of DBRC on CANDU lattice calculations.
Enabling DBRC for depletion calculations has a noticeable effect on
For burnup calculations, the fuel was split into four depletion regions, with one for each ring of pins. The burnup parameters shown in Tables
Serpent 2 burnup calculation parameters.
Parameter | Value |
---|---|
fpcut |
|
ttacut |
|
Serpent burnup calculation steps.
Step | Burnup (MWd/kg(U)) |
---|---|
|
0 |
|
0.03 |
|
0.06 |
⋯ | ⋯ |
|
0.21 |
|
0.25 |
|
0.30 |
|
0.45 |
|
0.60 |
⋯ | ⋯ |
|
1.20 |
|
1.50 |
|
1.80 |
|
2.10 |
|
2.50 |
|
3.00 |
|
3.50 |
⋯ | ⋯ |
|
6.00 |
|
7.00 |
|
8.00 |
|
9.00 |
|
10.00 |
⋯ | ⋯ |
|
30.00 |
The predictor-corrector method was enabled with linear interpolation and linear extrapolation to reduce the error in the computation of the irradiated nuclide densities. Unresolved resonance probability sampling was turned off (the default in Serpent) as its effect on thermal systems is relatively small. A bundle power of 700 kW was used for the burnup calculation. In all cases, the fuel temperature was assumed to be constant throughout the fuel cycle.
In addition, supplemental Serpent 2 runs were performed to obtain the exit burnup for each material as a function of enrichment. This allows the required enrichment for a specific burnup to be interpolated even for the cases which were not fully analyzed and allows for a better comparison of the different fuels in terms of enrichment.
For each case, the coolant void reactivity (CVR) and fuel temperature coefficient (FTC) were calculated for fresh fuel and for a burnup of 4 MWd/kg(U) or 96 MWh/kg(U). For the cases where the exit burnup exceeded 200 MWh/kg(U), an additional step closest to midburnup was selected from which to compute the CVR and FTC. For the CVR, the reactivity difference between the reference branch and coolant-voided branch was computed. The number of histories used was chosen to get sufficiently precise results. For the FTC, the fuel temperature was perturbed by 100 K in each direction. A greater number of neutron histories were used to improve the precision of the results.
To calculate the flux distribution for the temperature calculation, each pin was binned into bins equally spaced by radius, with five bins for non-FCM fuel and 20 bins for FCM fuel.
For non-FCM fuel, the following equation is fit to the flux distribution given by Serpent 2 along with the total power of 700 kW for the bundle, with
For FCM fuel, the power distribution is modelled directly as a piecewise function in FlexPDE. To determine the average pin temperature, the power for each bin is averaged amongst all 37 pins to get the power distribution. For the maximum pin temperature, the following multiplication is performed, where
Thermal conductivities and relations for mixtures were looked up in literature [
For the TRISO fuel, the properties shown in Table
Composition of TRISO particle [
Layer | Material | Outer radius | Vol. Frac. | Thermal conductivity |
---|---|---|---|---|
( |
(%) | (W m−1 K−1) | ||
Kernel | UN | 350 | 36.4 | Temperature dependent |
Buffer | Low-density PyC | 400 | 18.0 | 0.5 |
Inner PyC | High-density PyC | 435 | 15.6 | 4 |
SiC Coating | SiC | 470 | 18.3 | Temperature dependent |
Outer PyC | High-density PyC | 490 | 11.8 | 4 |
The final stage of the assessment involved modelling a simplified large break LOCA with loss of ECC. This can be considered to be the most serious credible loss of heat sink accident for CANDU, as the heat sink is lost almost immediately and the net-energy deposited in the first few seconds is significant. The model employed here is simplified as the thermal hydraulic circuit was not explicitly modelled in this study, but instead the thermal hydraulic data was modelled using information and data on large break LOCAs obtained from literature [
In summary, the thermal hydraulics involved a blowdown period where some cooling is sustained. After the initial power transient, the reactor would trip so that fission power would drop to near zero; thus the fuel would initially cool under these blowdown conditions. After the blowdown period the coolant would stagnate and the convection of heat to the coolant would drop to nearly zero, so the fuel would heat up due to decay heat. Radiation of heat between the rings of pins and the pressure tube was modelled during this phase, with the pressure tube deformation to the calandria assumed to occur at 1200 K.
The reactivity feedback due to coolant voiding, based on a two-loop, two-pass CANDU design, was
The core reactivity with all feedback was modelled by
For each case, the transient was modelled in two steps. The first step only included the “average” pin and modelled the point kinetics with one delayed group. The second step modelled the pins of the four rings along with the pressure tube, using the fission power history calculated in the first step.
The progression of the loss of coolant in the model is given in Figure
HTS pressure and void fraction behaviour for stylized transient.
Certain ATF selections required a greater enrichment than other cases. As silicon carbide captures very few neutrons; the UO2-SiC composite fuels only require minimal enrichment to provide the same exit burnup. U-9Mo requires slight enrichment (0.93%) due to the neutrons captured by molybdenum isotopes. For uranium nitride, if natural nitrogen were used, a significant enrichment of the uranium would be required. However, by enriching the nitrogen instead, natural uranium can be used and the exit burnup exceeds the 200 MWh/kg(U) for UO2 bundles. As the uranium density is greater for the UN bundles, the 4044 MWh/bundle of UO2 bundles is exceeded by UN bundles by enriching only the nitrogen. For FCM fuel, the enrichment requirements are increased due to the replacement of fissile material with inert material. However, this effect becomes stronger when the exit burnup is increased to compensate for the low uranium density of the fuel. Reducing the packing fraction increases enrichment requirements further due to reduction of the uranium density. As UN kernels contain more uranium than UO2 kernels, the enrichment requirements are less for UN kernels than UO2 kernels. The TRISO fuel has the lowest uranium mass and therefore requires the highest enrichment to achieve a similar bundle lifetime.
Table
Required enrichment for ATF.
Fuel |
Enrichment (wt% 235U) for target exit burnup |
||
---|---|---|---|
200 MWh/kgU | 600 MWh/kgU | 4044 MWh/bundle | |
UO2 |
|
1.30% |
|
UO2 + 3% SiC (zirc4 clad) |
|
1.32% |
|
UO2 + 3% SiC (SiC clad) |
|
|
|
UO2 + 3% SiC (zirc4 + FeCrAl clad) |
|
|
|
UO2 + 6% SiC | 0.72% | ||
UO2 + 10% SiC |
|
|
|
UN |
|
1.22% | |
UN + 3% U3Si2 | <0.711%; NU = 231.2 MWh/kgU | ||
UN + 6% U3Si2 | <0.711%; NU = 230.6 MWh/kgU | ||
UN + 10% U3Si2 |
|
||
UN + 10% ZrN | <0.711%; NU = 210.4 MWh/kgU | ||
U-9Mo |
|
|
|
UO2 FCM 35% packing | 1.07% | 1.89% | |
UO2 FCM 40% packing |
1.01% | 1.80% |
|
UO2 FCM 45% packing | 0.95% | 1.43% | |
UO2 FCM 50% packing | 0.91% | 1.28% | |
UO2 FCM 55% packing | 0.88% |
|
1.16% |
UN FCM 35% packing | 0.93% | 1.28% | |
UN FCM 40% packing | 0.89% |
|
|
UN FCM 45% packing | 0.85% | 1.02% | |
UN FCM 50% packing | 0.82% | 0.93% | |
UN FCM 55% packing | 0.80% | 0.87% | |
UN TRISO 55% packing | 1.12% |
|
Enrichment requirements for UO2 and ATF cases.
Enrichment requirements for FCM fuel for 4044 MWh/bundle exit burnup.
In all cases, the fuel temperatures calculated for the ATF fuel pins are less than those for the UO2 fuel pins, as shown in Figure
Estimated temperature of fresh and midburnup UO2-SiC fuel.
Fuel | Fresh fuel temperature (K) | Midburnup temperature |
---|---|---|
UO2 | 2070 | 2306 |
UO2 + 10% SiC | 1464 | 1881 |
Difference (K) | 606 | 425 |
Difference (%) | 29% | 18% |
Fuel temperatures calculated for ATF cases.
Fuel | Enrich. | Outer ring temperature (K) | Average pin temperature (K) | ||||
---|---|---|---|---|---|---|---|
Power = 880 kW | Power = 700 kW | ||||||
Fuel CL | Fuel Avg. | Clad | Fuel CL | Fuel Avg. | Clad | ||
UO2 | 0.711% | 2069.6 | 1248.4 | 608.3 | 1437.0 | 981.6 | 593.9 |
1.30% | 2159.0 | 1290.1 | 610.5 | 1429.1 | 979.4 | 593.9 | |
|
|||||||
UO2 + 3% SiC |
0.72% | 1760.3 | 1124.4 | 608.2 | 1269.7 | 913.5 | 593.9 |
1.32% | 1835.4 | 1157.7 | 610.4 | 1264.0 | 911.8 | 593.9 | |
|
|||||||
UO2 + 3% SiC |
0.711% | 1704.9 | 1087.0 | 596.6 | 1239.4 | 890.8 | 585.7 |
|
|||||||
UO2 + 3% SiC |
0.79% | 1779.2 | 1132.1 | 611.9 (zirc4) |
1269.0 | 913.0 | 596.2 (zirc4) |
|
|||||||
UO2 + 6% SiC | 0.72% | 1607.6 | 1064.4 | 608.0 | 1190.9 | 880.6 | 593.9 |
|
|||||||
UO2 + 10% SiC | 0.73% | 1463.7 | 1007.1 | 607.9 | 1115.5 | 848.5 | 593.9 |
|
|||||||
UN | 0.711% | 868.5 | 760.8 | 610.1 | 772.1 | 697.2 | 593.9 |
1.22% | 881.4 | 769.9 | 612.6 | 770.9 | 696.8 | 593.9 | |
|
|||||||
UN + 3% U3Si2 | 0.711% | 870.9 | 762.1 | 610.1 | 773.8 | 698.1 | 593.9 |
|
|||||||
UN + 6% U3Si2 | 0.711% | 873.0 | 763.2 | 610.0 | 775.7 | 699.1 | 593.9 |
|
|||||||
UN + 10% U3Si2 | 0.711% | 876.2 | 764.9 | 610.0 | 778.3 | 700.5 | 593.9 |
|
|||||||
UN + 10% ZrN | 0.711% | 856.9 | 753.8 | 609.3 | 767.0 | 694.4 | 593.9 |
|
|||||||
U-9Mo | 0.93% | 825.1 | 742.3 | 612.1 | 740.6 | 682.3 | 593.9 |
|
|||||||
UO2 FCM 35% packing | 1.07% | 814.8 | 727.0 | 605.9 | 744.2 | 681.5 | 593.9 |
1.89% | 821.5 | 731.5 | 607.1 | 743.9 | 681.4 | 593.9 | |
|
|||||||
UO2 FCM 40% packing | 1.01% | 836.0 | 737.3 | 606.1 | 758.5 | 688.4 | 593.9 |
1.62% | 842.3 | 741.4 | 607.1 | 758.1 | 688.3 | 593.9 | |
1.80% | 843.7 | 742.4 | 607.3 | 758.0 | 688.2 | 593.9 | |
|
|||||||
UO2 FCM 45% packing | 0.95% | 862.1 | 749.8 | 606.2 | 775.8 | 696.6 | 593.9 |
1.43% | 867.9 | 753.5 | 607.1 | 775.4 | 696.5 | 593.9 | |
|
|||||||
UO2 FCM 50% packing | 0.91% | 893.7 | 765.0 | 606.4 | 796.7 | 706.6 | 593.9 |
1.28% | 899.2 | 768.4 | 607.2 | 796.3 | 706.5 | 593.9 | |
|
|||||||
UO2 FCM 55% packing | 0.88% | 933.4 | 784.0 | 606.6 | 822.7 | 719.1 | 593.9 |
1.16% | 938.7 | 787.3 | 607.3 | 822.3 | 719.0 | 593.9 | |
|
|||||||
UN FCM 35% packing | 0.93% | 772.7 | 707.6 | 606.7 | 712.9 | 666.4 | 593.9 |
1.28% | 775.9 | 709.8 | 607.4 | 712.8 | 666.3 | 593.9 | |
|
|||||||
UN FCM 40% packing | 0.89% | 779.6 | 711.1 | 607.0 | 717.0 | 668.4 | 593.9 |
1.13% | 782.1 | 712.9 | 607.5 | 716.9 | 668.3 | 593.9 | |
|
|||||||
UN FCM 45% packing | 0.85% | 786.4 | 714.7 | 607.2 | 721.2 | 670.4 | 593.9 |
1.02% | 788.3 | 716.1 | 607.7 | 721.1 | 670.3 | 593.9 | |
|
|||||||
UN FCM 50% packing | 0.82% | 793.1 | 718.3 | 607.5 | 725.3 | 672.4 | 593.9 |
0.93% | 794.5 | 719.3 | 607.8 | 725.2 | 672.4 | 593.9 | |
|
|||||||
UN FCM 55% packing | 0.80% | 799.3 | 721.8 | 607.7 | 729.1 | 674.4 | 593.9 |
0.97% | 800.5 | 722.6 | 608.0 | 729.0 | 674.4 | 593.9 | |
|
|||||||
UN TRISO 55% packing | 2.25% | 988.0 | 813.7 | 607.1 | 861.6 | 739.5 | 593.9 |
Fuel temperatures for UO2 and ATF cases.
Overall, there are moderate differences between the different cases in terms of coolant void reactivity and fuel temperature coefficient, as seen in Figure
Coolant void reactivity and fuel temperature coefficient for ATF.
Description | Enrichment |
Fresh fuel |
Fresh fuel | Midburnup | ||
---|---|---|---|---|---|---|
CVR | FTC | CVR | FTC | |||
(mk) | (pcm/K) | (mk) | (pcm/K) | |||
UO2 | 0.711% | 1.11850 |
|
|
|
|
1.300% | 1.35036 |
|
|
|
|
|
|
||||||
UO2 + 3% SiC | 0.720% | 1.12319 |
|
|
|
|
1.320% | 1.35597 |
|
|
|
|
|
|
||||||
UO2 + 6% SiC | 0.720% | 1.12203 |
|
|
|
|
|
||||||
UO2 + 10% SiC | 0.730% | 1.12577 |
|
|
|
|
|
||||||
UN | 0.711% | 1.11918 |
|
|
|
|
1.220% | 1.32137 |
|
|
|
|
|
|
||||||
UN + 3% U3Si2 | 0.711% | 1.11907 |
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UN + 6% U3Si2 | 0.711% | 1.11929 |
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UN + 10% U3Si2 | 0.711% | 1.11905 |
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UN + 10% ZrN | 0.711% | 1.11744 |
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U-9Mo | 0.930% | 1.12952 |
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UO2 FCM in SiC |
1.070% | 1.18147 |
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1.890% | 1.41177 |
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UO2 FCM in SiC |
1.010% | 1.17808 |
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1.620% | 1.37165 |
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1.800% | 1.41008 |
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UO2 FCM in SiC |
0.950% | 1.16881 |
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1.430% | 1.33787 |
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UO2 FCM in SiC |
0.910% | 1.16434 |
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1.280% | 1.30639 |
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UO2 FCM in SiC |
0.880% | 1.16149 |
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1.160% | 1.27726 |
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UO2 FCM (650 |
1.010% | 1.17804 |
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1.620% | 1.37161 |
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UO2 FCM (750 |
1.010% | 1.17809 |
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1.620% | 1.37170 |
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UN FCM in SiC |
0.930% | 1.16290 |
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1.280% | 1.29615 |
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UN FCM in SiC |
0.890% | 1.15883 |
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1.130% | 1.25901 |
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UN FCM in SiC |
0.850% | 1.15066 |
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1.020% | 1.22738 |
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UN FCM in SiC |
0.820% | 1.14447 |
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0.930% | 1.19745 |
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UN FCM in SiC |
0.800% | 1.14123 |
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0.970% | 1.17653 |
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UN FCM in SiC |
2.250% | 1.46143 |
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UO2 + 3% SiC |
0.720% | 1.12319 |
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UO2 + 3% SiC |
0.711% | 1.12010 |
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UO2 + 3% SiC |
0.790% | 1.12855 |
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CVR and FTC for UO2 and ATF cases.
The results of the stylized transient were found to be similar irrespective of ATF design. As a result of the positive coolant void reactivity for CANDU, power increases fairly rapidly until the transient is turned over by negative reactivity inserted by the reactor’s shutdown systems [
Reactor power transient for UO2 and ATF.
The postblowdown heat-up phase is governed primarily by the heat capacity of the fuel and cladding material, distribution of decay power, and thermal radiation between the fuel rings and pressure tube. Initially, the outer ring is at the highest temperature due to a higher decay power, particularly for more strongly absorbing fuel where the spatial self-shielding effect is stronger. However, at higher temperatures, this becomes inverted, with the centre pin being hottest, as the other three rings insulate one another and the centre pin. In all cases, however, the peak temperature is in a narrow range between 1800 K and 2000 K. Therefore, in terms of fuel temperature behaviour, the choice of fuel material does not have a substantial effect on the progression of an accident. However, since U-9Mo has a melting point of only around 1400 K, fuel melting will occur for this fuel material before the pressure tube can sag and the moderator can be established as a heat sink. Therefore, in terms of progression of a LOCA-LOECC accident, or similar degraded-cooling scenarios such as a SBO, ATF options do not provide a substantial advantage over UO2 in delaying severe consequences, as the progression of these accidents is governed by large nonfuel heat sinks, such as the moderator and shield tank water. The U-9Mo case in particular is inferior due to the fuel melting which occurs much sooner than for UO2.
Two of the cases tested are shown in Figure
Fuel temperature behaviour for selected cases: UO2 (a) and U-9Mo (b).
The most modest change from standard UO2 fuel in zircaloy cladding is to create a UO2-SiC composite fuel. Overall, the addition of SiC adds negligible parasitic absorption, but the displacement of UO2 must be compensated by a very slight enrichment in order to achieve the same exit burnup, and slightly more to achieve the same bundle energy. However, the temperature reduction is significant for just 10% of the fuel volume being SiC, as the centre-line temperature is reduced by about 30% for fresh fuel, or about 20% for irradiated fuel. In terms of physics parameters, there is a small but significant reduction in CVR, with only a slight effect on FTC.
Uranium nitride has the potential downside of natural nitrogen being a strong neutron absorber, requiring significant uranium enrichment to compensate and reducing the uranium economy of the fuel cycle. To avoid this, nitrogen enriched in
Adding U3Si2 to UN reduces the exit burnup only slightly. Adding ZrN to make a
Uranium-molybdenum fuel also requires enrichment (0.93% for the same exit burnup) as molybdenum also contributes to parasitic absorption, though not as severely as nitrogen. However, as the uranium density is significantly greater than it is for UO2, the fuel lasts longer in the core for this same exit burnup. Also, the fuel temperature coefficient remains significantly negative even with irradiation. However, U-9Mo fuel has several disadvantages. A greater coolant void reactivity results in a larger power pulse in a LOCA event. More importantly, though, the melting point is low enough that the fuel can melt in a number of accident scenarios for which UO2 fuel or other ATF candidates would easily survive.
Fully ceramic microencapsulated fuel, in terms of uranium density, is the opposite of UN and U-9Mo fuel, as the uranium density is lower for bare UO2 or UN kernels in a SiC matrix compared to solid UO2 and is even lower for true FCM fuel where TRISO particles are embedded in a SiC matrix. Therefore, even though the SiC matrix contributes to very little parasitic absorption, significant enrichment is required to achieve the same energy per bundle as a UO2 bundle. The CVR is significantly less for FCM fuel compared to UO2 fuel, with a lower uranium density corresponding to a lower CVR. However, the FTC for FCM fuel is generally slightly less negative at midburnup compared to the corresponding solid fuel and is even slightly positive for some cases with UO2 kernels. The effect on FCM is a combination of a positive contribution from the material change and a negative contribution from the reduced fuel temperature.
Changing the cladding also has a small effect on reactor physics. Silicon carbide has a beneficial effect as it is more transparent to neutrons than zircaloy; thus the fuel enrichment can be reduced by about 0.01%. On the other hand, FeCrAl coating requires roughly a 0.07% increase in enrichment for the case it was tested on, despite the fact that only 20% of the cladding thickness was replaced. Small but significant reductions in CVR were noted, while changes in FTC were insignificant.
Given the above, in the transient scenario modelled no substantial advantage can be identified for any of the ATF candidates over UO2. The progression of a severe accident in a CANDU reactor is largely governed by several large heat sinks, including the moderator and shield tank water. These heat sinks, along with structural materials such as pressure tubes, are independent of any changes to the fuel materials, and they have a large heat capacity compared to the fuel itself. Since decay heat is governed by the reactor’s operating power history and energy deposition in a transient and not by the fuel composition, the accident is expected to progress at a similar rate regardless of the fuel composition. The primary exception is for U-9Mo fuel, which is actually inferior to UO2 fuel due to premature melting in a number of scenarios.
Despite this, there are a few places where, qualitatively, ATF can be expected to have an advantage over UO2. The first advantage is in terms of fission product retention within fuel pellets. Lowering the fuel temperature is expected to reduce the diffusion rate of certain fission products into the fuel-clad gap. In the case of fuel damage where the cladding fails, only these fission products escape, while those trapped within the fuel pellets are retained. In addition, FCM-type fuels provide additional barriers to fission product release. However, in all cases, including for UO2, the radiation release due to cladding failure is not expected to be substantial.
The second advantage applies to ATF cladding, for which a reduction in hydrogen generation is expected under certain accident conditions compared to zircaloy, reducing the risk of further failures leading to a hydrogen explosion. SiC cladding can potentially withstand high temperatures without significant oxidation. By contrast, however, FeCrAl has a lower melting point than zircaloy and would actually melt under the stylized accident scenario that was modelled, though zircaloy oxidation would still be delayed compared to not having the coating.
The primary disadvantage for ATF over the current standard UO2 fuel is the need for enrichment. Even for a UO2-SiC composite bundle with 10% SiC, while it can be used without enrichment, the energy that can be extracted from one fuel bundle is reduced by about 25%, which has economic implications. U-9Mo and FCM result in a subcritical or barely critical core. Applying a FeCrAl surface layer to the cladding also substantially reduces exit burnup unless the fuel is enriched. Even the very slight enrichment predicted for UO2-SiC composite bundles would require the establishment of new infrastructure, substantially raising costs even if relatively little enrichment is required. This is in contrast to ATF considerations for LWRs, where an increase in enrichment from, for example, 4.8% to 5.0%, would not require significant infrastructure changes. Economically, considering ATF for CANDU reactors is more reasonable in the case where the fuel input is not natural uranium, such as if a CANDU is being fueled by recycled uranium (from used LWR fuel), MOX, thorium, or any combination of those with each other or with natural or depleted uranium. In this case, the slight enrichments required for ATF can be accommodated more easily. However, the reactor physics and heat transfer properties would have to be reevaluated for the fuel change.
There are two primary exceptions. The first exception is for the UN-based fuel materials, when enriched in
More generally, changes to the fuel geometry were not considered in this study. One possibility that could affect the fuel properties would be changing to 43-element CANFLEX fuel or to some other geometry with different pin sizes. ATF materials improve the safety margin over UO2 for centre-line temperature, potentially permitting the use of bundles with fewer, larger pins to increase the fuel volume. However, such a change would reduce the fuel surface area and thus reduce the margin to critical heat flux (CHF) and so cannot be done if CHF is the limiting factor. Alternatively, if centre-line temperature is the limiting factor, the maximum bundle power and possibly the reactor power can be uprated. Since CHF for high-density fuel is a potential issue, that fuel could potentially benefit from geometry with smaller pins for the outer ring, such as the CANFLEX bundle or possibly a fuel-specific bundle with even greater differences in pin diameters. Another possibility would be to use graded enrichments, with the fuel in the outer ring having a lesser enrichment than the rest of the fuel.
The primary findings of this study are as follows: UO2 with SiC cladding, along with UN fuel enriched in In a number of severe accident scenarios that were considered, changing the fuel to ATF would not extend the time between the onset of the accident and the onset of fuel melting. The propagation of the accident generally allows the moderator, and possibly the shield tank water, to act as a heat sink for the fuel’s decay heat to prevent melting. The size of these heat sinks and the amount of decay heat are independent of the fuel material, and as these heat sinks are large relative to the heat capacity of the fuel, changing the fuel has relatively little effect on when melting will occur. The key exception is U-9Mo, which has a melting point of roughly 1400 K and so heat transfer to the moderator is not sufficient to prevent melting, making this material accident- All ATF fuel pellet materials reduce the fuel temperature compared to UO2 and can potentially reduce the diffusion rate of fission products into the fuel-clad gap. This can reduce the magnitude of radiation release in the case of cladding failure. However, this effect was not quantified in this study, and the significance of this advantage depends on whether the predicted releases from UO2 fuel under cladding failure (no fuel melting) can be accepted. In an accident scenario, it is predicted that cladding temperatures are hot enough to result in extensive oxidation. Therefore, oxidation-resistant cladding can delay or prevent the oxidation of cladding and production of hydrogen gas. Again, evaluation of this effect was beyond the scope of the models used in this study, but the results of this study show that it would be useful for future studies to evaluate the extent of cladding oxidation and hydrogen generation, and the comparative risk of a hydrogen explosion for each case.
Further studies are needed to evaluate the full-core behaviour of the fuel, possible changes to fuel geometry, and the mechanical behaviour of the fuel and cladding materials in normal and accident conditions. In addition, a more detailed economic analysis should be carried out to determine whether it is viable to switch to accident-tolerant fuel, weighing increased costs due to enrichment and fuel manufacturing against the reduced level of risk and also weighing the costs against the costs of other modifications that can be made to the station to reduce the risk of severe accident consequences.
While some cases, such as U-9Mo, can be immediately ruled out from consideration, additional cases not analyzed in this study could be considered. This can include fuel with other “dopant” materials mixed in to improve thermal conductivity. It can also go beyond simple material changes, such as changing the bundle geometry (e.g., to CANFLEX) or using different compositions for different rings, which can improve bundle power distribution but increase the complexity of the bundle for manufacture.
In most cases, an economic disadvantage is probable as either uranium enrichment is needed or the fuel otherwise becomes more expensive to manufacture (which can be expected for silicon carbide cladding). Uranium nitride enriched in
The authors declare no conflicts of interest regarding the publication of this paper.