which are able to enhance radiation-resistance of metal uranium fuel, and, finally, casting of uranium rods.
Some data on physical properties of uranium-based fuels (density, melting temperature, heat conductivity and volumetric swelling) are presented in Table 1.5.
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Table 1.5 |
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Physical properties of uranium-based fuels |
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Fuel type |
Density, |
Melting |
Heat conductivity, |
Swelling |
g/cm3 |
temperature, оС |
W/m·К |
∆V/V, % |
|
Metal |
18,67 |
1130 |
28 (27оС) |
4-6 |
44 (727оС) |
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UO2 |
10,96 |
2780 |
8,4 (45оС) |
1,4-1,5 |
2,4 (1327оС) |
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UC |
13,63 |
2350 |
32,7 (45оС) |
1,7-1,8 |
7,3 (500оС) |
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UN |
14,32 |
2650 |
16 (200оС) |
1,5-1,6 |
21 (800оС) |
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1.3.2. Fabrication of fuel rods and fuel assemblies
The following requirements must be satisfied by the manufacturing technologies of fuel rods and fuel assemblies:
1.Designs of fuel rods and fuel assemblies, physical properties of fuel and structural materials must be able to ensure long-term mechanical strength, stability of forms and sizes during a reactor lifetime.
2.Materials of fuel rods (fuel meat, cladding, fuel-cladding gap) must be chemically compatible and mutually stable, i.e. any fuel-cladding interactions that can cause radiation embrittlement and plasticity loss must be excluded.
3.The cladding materials must be insoluble and corrosion-resistant in cladding-coolant interactions.
4.Structural materials of fuel rods and fuel assemblies must be sufficiently weak neutron absorbers (minimal cross-sections of neutron radiative capture).
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5.Designs and the manufacturing technologies of fuel rods and fuel assemblies must exclude any possibilities of local overheating. This exclusion can be achieved by undertaking the following countermeasures:
a. Uniform distribution of fissile isotopes in fuel rods.
b. Availability of the contact interlayer (fuel-cladding gap) filled up with helium or sodium for intensification of heat removal processes and for prevention of fuel-cladding interactions.
c. Strict spatial separation of fuel rods from each other by special spacers with proper accounting for potential shortening of the inter-rod gap in the process of the reactor operation.
6.Designs and the manufacturing technologies of fuel rods and fuel assemblies must be sufficiently simple for their mass production.
7.Designs, the manufacturing technologies and selection of materials for fuel rods and fuel assemblies must take into account a feasibility of sufficiently simple dismantling procedures for spent fuel reprocessing.
The manufacturing process of UO2-based fuel rods and fuel assemblies includes the following stages:
1.Preparation of nuclear fuel (conversion of uranium hexafluoride into uranium dioxide powder, granulation and sintering of fuel pellets).
2.Preparation of fuel cladding (flaw detection and quality control).
3.Preparation of the completing details for mounting of fuel assemblies (wrappers, end caps, spacers).
4.Manufacturing of fuel rods: insertion of fuel pellets into tubular claddings, installation of end caps, filling up with helium (as a fuel-cladding gap and a flaw detector thanks to high permeability of helium), sealing of fuel rods by welding, quality control.
Recently, the Russian R&D Institute of Atomic Reactors in Dimitrovgrad has developed the fuel rod manufacturing technology that constitutes an alternative to the technology based on fabrication of fuel pellets. UOXor MOX-fuel granules are used as the feed material. Some amount (up to 5% HM) of metal natural uranium powder (so called getter) is blended with other fuel granules. Main mission of the getter consists in absorption of oxygen atoms released from fuel particles in fission reactions. As a result, oxygen atoms can not move to the fuel cladding, their corrosion activity is neutralized. The fuel-getter blend is introduced into tubular cladding and packed by vibration. Sufficiently
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high fuel density (above 90% of its theoretical value) can be achieved. The experiences gained during pilot usage of the vibration-compacted fuel rods in the research fast reactor BOR-60 demonstrated a high efficiency of such technology. The record values of fuel burn-up were reached (about 32%HM) with specific volumetric swelling at the level of 0,6% V/V per one percent of fuel burn-up. For comparison, metal uranium fuel and UOX-fuel can swell up with the rate of 4-6% V/V and 1,4-1,5% V/V, respectively, per one percent of fuel burn-up.
5. Assemblage of fuel rods into a single fuel assembly, quality control and testing.
The following concluding remarks can be made on the manufacturing technologies of fuel rods and fuel assemblies:
a.The manufacturing technology is a mass production and highly automated process.
b.The manufacturing technology is a high-precision process.
The manufacturing process of RBMK-1000 core requires about 200 thousand completing details, 14 million fuel pellets and 240 thousand welded joints.
All the manufacturing procedures must be put under strict quality control with application of the computer-aided NM control and accountability network. The nuclear fuel fabrication plant is a very significant area for functioning of a reliable and highly effective NM physical protection, control and accountability (MPC&A) system.
In the future, when nuclear fuel cycle will be closed, even fresh nuclear fuel is characterized by intense radioactivity and residual heat generation. This will require applying only the newest remote technologies for manufacturing of fuel rods and fuel assemblies. Accordingly, all components of MPC&A system become more complicated and must be more sophisticated.
1.4. Use of nuclear fuel in nuclear power reactors
1.4.1. Purposes of nuclear reactor refueling
The major NFC stage is an energy utilization of nuclear fuel in nuclear power reactors.
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In order to provide sufficiently long operation of nuclear power reactors (up to 18 months as in some advanced LWR), appropriate amount of nuclear fuel must be loaded and properly arranged in the reactor core. Neutron-physical properties of the loaded fuel must ensure large enough reactivity margin to compensate negative reactivity effects caused by depletion of fissile isotopes and build-up of fission products, parasitic neutron absorbers. This means that, before the reactor operation starts up, the reactor is essentially supercritical but the reactivity margin (КEFF - 1) must be suppressed by the regulatory mechanisms: control rods made of natural (or enriched with 10B isotope, very strong neutron absorber) boron carbide, boric acid dissolved in light-water coolant, burnable poisons (gadolinium, erbium) introduced into fuel compositions.
As content of fissile isotopes in fuel decreases and content of fission products (parasitic neutron absorbers) in fuel increases, the moment comes when the reactor can not be longer critical. All control rods are already withdrawn from the reactor core, all boric acid is removed from light-water coolant. Nevertheless, the reactor becomes sub-critical (КEFF < 1).
For the reactor operation to continue, NPP operator has to undertake some corrective measures to return the reactor to the supercritical state (КEFF > 1). The following actions can be performed:
1.Full or partial substitution of fresh fuel assemblies for irradiated ones.
2.Partial transpositions (shuffling) of irradiated fuel assemblies from one region of the reactor core to another.
3.Any combinations of two aforementioned actions.
A set of these actions undertaken to restore the reactor ability for
long-term operation is named the refueling strategy.
Thus, the first and major mission of the refueling is to replenish the reactivity margin and enable the reactor to continue a long enough operation at nominal power level during a certain time interval.
The second, also important but supplementary, mission of the refueling is to ensure as flat as possible spatial shape of heat generation rate. If spatial distribution of heat generation rate in the reactor core is uniform enough, then all fuel assemblies are used under maximal acceptable level of energy production. So, maximal value of total energy out-
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put can be obtained from the reactor core. In addition, uniform spatial distribution of heat generation rate ensures uniform fuel burn-up, i.e. all fuel assemblies discharged from the reactor core are characterized by the same (or nearly the same) isotope compositions.
However, these benefits (maximal energy output and identical isotope compositions of spent fuel) can not be obtained in a uniformly fueled reactor. In such reactors heat generation rate is highest in the core center and drops down nearly to zero at the core periphery. The average value of heat generation rate is approximately three times lower than its maximal value in the core center. Non-uniform fuel loading is required to flatten spatial shape of heat generation rate and fuel burn-up. For example, uranium fuel of the lower enrichment may be placed in central region of the reactor core while uranium fuel with relatively higher enrichment may be placed at the core periphery. As a rule, main control rods are also placed in central region of the reactor core. That is why neutron flux and heat generation rate in the core center are lower than those at the core periphery.
Thus, formation of such a fuel loading that is characterized by as flat as possible spatial shape of heat generation rate in the reactor core constitutes a supplementary but very important mission of the refueling strategies.
1.4.2. Strategies of nuclear reactor refueling
There are many various refueling strategies in nuclear power reactors.
1. The simplest refueling strategy presumes a uniform fuel loading with its complete removal and replacement with fresh fuel loading when the reactor criticality can not be longer maintained. This refueling scheme is called the “batch irradiation” and not used now because of the following serious drawbacks:
a.Spatial shape of heat generation rate in the reactor core is quite non-uniform with the peaking factor (peak-to-average ratio) above three.
b.Content of fissile isotopes decreases more intensely in the central core region than at the core periphery. So, very uneven fuel burn-up takes place in the uniformly fueled reactor core. After each irradiation
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