of coolant flow which alternates between adjacent fuel channels. So, the refueling process in two adjacent channels is done in two mutually opposite directions. The insertion of fresh fuel bundles into peripheral regions of the reactor core from two opposite sides can upgrade heat generation rate at the reactor periphery and, thus, flatten spatial distribution of heat generation rate.
Russian RBMK-type reactors (light-water cooled, graphite moderated reactors, LWGR) can be also refueled in a continuous on-line manner, like CANDU-type reactors because both reactor types are channel reactors that made it possible to arrange refueling of any fuel channel individually. A dedicated loading-unloading machine (LUM) containing one fresh fuel assembly and space for disposition of one spent fuel assembly can do the following refueling operations:
1.The LUM filled up with the warm condensate (300С) attaches onto the fuel channel to be refueled.
2.Pressure in the fuel channel and pressure in the LUM cask used for disposition of spent fuel assembly are equalized (~75 atmospheres).
3.The fuel channel and the LUM form a single circuit. The warm condensate is pumped into the circuit.
4.Spent fuel assembly is grappled by the LUM manipulator and withdrawn from the fuel channel.
5.Passability of the fuel channel is checked up with a fuel assembly imitator.
6.Fresh fuel assembly is inserted into the fuel channel.
7.The fuel channel is locked, pressure in the LUM decreases to the ambient level, the LUM and the fuel channel are disconnected.
It is evident that those nuclear power reactors which can be refueled continuously, without the reactor shutdown for several weeks, i.e. CANDU and RBMK reactors represent a particular threat to nuclear non-proliferation regime because of the following reasons:
1.Operators of CANDU and RBMK reactors are able, in principle, to conduct relatively short-term (two-three months) irradiation of uranium fuel assemblies for unauthorized build-up of weapon-grade plutonium.
2.To prevent the unauthorized short-term irradiations of uranium fuel assemblies, a continuous (not periodical) presence of the IAEA inspectors is required at the operating CANDU and RBMK reactors.
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3. All the MPC&A-related measures are more difficult for undertaking at the reactors with continuous refueling operation mode.
Spent fuel assemblies are intense radiation and decay heat sources. That is why all NPP reactors are provided with spent fuel water pools where spent fuel assemblies are stored until their radioactivity and residual heat generation rate drop below the acceptable levels for transportation, reprocessing or ultimate disposal.
The spent fuel storage pools must be equipped with the following auxiliary systems:
1.Residual heat removal system.
2.Ion-exchange installation for water purification and removal of solid radioactive particles.
3.Ventilation installation for air purification and retention of gaseous radioactive materials by special super-filters.
Interim storage of LWR spent fuel assemblies in NPP water pools can last up to 10 years. The same storage time is chosen for LMFBR spent fuel assemblies. Previously, it was thought that the backend part of the closed NFC including LMFBR spent fuel reprocessing and plutonium recycle must be as short as possible (6-12 months as a target value) to supply the developing nuclear power industry with plutoniumbased fuel. Nowadays, there are no imperative reasons for the worldwide deployment of LMFBR-based NPP. So, the same storage time was adopted for spent fuel assemblies discharged from LWRand LMFBRtype reactors.
Upon the expiry of the interim storage time, spent fuel assemblies can be transported to the deep underground repositories for ultimate disposal (the open NFC option) or to the spent fuel reprocessing plants for plutonium recovery and energy utilization (the closed NFC option). Relatively long interim storage time and weak development of nuclear technologies intended for ultimate disposal or radiochemical reprocessing of spent fuel assemblies resulted in gradual exhaustion of the water pools capacity. To neutralize these negative effects, the following countermeasures are undertaken:
1.The tighter positioning of spent fuel assemblies inside the water pools under the stricter nuclear safety control.
2. Partitioning of the water pools by metal structures containing strong neutron absorbers (boron, for instance).
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3. Build-up of the centralized large spent fuel storages.
1.4.4. Transportation of spent nuclear fuel
Transportation is a necessary link between all NFC stages, and is especially significant for transportation of spent fuel assemblies. Spent fuel assemblies may be shipped by all transport means (trucks, railroad, river boats or sea-going ships) except of aircrafts. According to the RF regulations, all shipments of nuclear materials with specific radioactivity above 2 µCi/kg are regarded as radiation shipments. Specific radioactivity of spent fuel equals about 1 MCi/kg.
The spent fuel transport casks can weigh about 100 tons. Total weight of spent fuel assemblies in the transport casks is about 2-5% from total weight of the transport cask. The remaining 95-98% of total weight belongs to the transportation safety systems.
A typical spent fuel transport cask looks as follows:
1.Large hollow thick-walled cylinder in a vertical or horizontal (preferentially) position (1,5-2 m in diameter, 4-6 m in length, wall is about
40cm thick) made of steel, cast iron or concrete.
2.Outer surface of the transport cask is covered by special fins for ex-
tension of the heat removal area ( 30 m2). The finned outer surface extends the heat removal area approximately twice.
3.Inner surface of the transport cask is lined with stainless steel to enhance corrosion-resistance. The inner liners can include some layers of neutron moderators and neutron absorbers (borated polyethylene, for instance).
4.Metal shelves for disposition of spent fuel assemblies are placed in the inner cavity of the transport cask. During shipment, the inner cavity is filled up with coolant. Decay heat is removed from spent fuel assemblies either by natural convection or forced circulation depending on the value of heat generation rate.
5.The transport casks are hermetized with application of the reinforced densifiers.
6.The transport casks are equipped with control systems for permanent monitoring of the inner cavity parameters (radioactivity, decay heat generation rate, temperature and pressure of coolant) and with accidental decontamination system.
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The following requirements must be satisfied by designers of the spent fuel transport casks:
1.Reliable radiation protection of the staff involved, population and the environment against neutron and gamma-emissions (metal vessel containing high-efficiency neutron moderators and neutron absorbers).
2.Reliable nuclear safety ensuring (metal shelves containing strong neutron absorbers, limitations on the number of spent fuel assemblies to be loaded into the transport casks).
3.Reliable removal of decay heat (the finned outer surface, forced circulation of coolant in the inner cavity of the transport cask).
4.Reliable hermetization of the transport casks even under severe accidental conditions.
To check up hermeticity of the transport casks, they must undergo the following severe tests:
a. Drop test from 9-m height onto a steel plate.
b. Puncture test from 1-m height onto a vertical metal rod (15 cm in diameter).
c. Immersion test in light water (depth - 15 m, duration – 8 hours).
d. Fire test – staying in flame at 800 0С for 30 minutes plus 2-hour staying without a forced cooldown.
The IAEA has elaborated the following requirements to thermal pa-
rameters of the transport casks in operation:
a. Temperature of the cask surface must be below 820С at the ambient air temperature of 380С.
b. Internal coolant pressure in the cask cavity must be below 7 atmospheres.
c. The outer surface of the cask must be extended by fins. Additional 30 m2 of the heat removal area would be large enough for safe shipment of
spent fuel assemblies with total power from 25 to 30 kW ( 30 spent fuel assemblies from VVER-440 after 3-year cooling period) without a forced cooldown.
1.5. Radiochemical reprocessing of spent nuclear fuel
The following aims are pursued by technologies intended for spent nuclear fuel (SNF) reprocessing:
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1.Recovery of accumulated plutonium and residual uranium for the repeated use (recycle) as fissile and fertile materials.
2.Separation of fission products and transuranium elements for further treatment as radioactive wastes.
The IAEA has worked out the following recommendations on the reprocessing quality of spent fuel assemblies discharged from power LWR after the standard operation cycle (fuel burn-up - 33 GWd/t, the cooling time – 10 years):
1.Extent of plutonium and uranium recovery – above 99 ,9%.
2.Decontamination factors, i.e. ratios of an impurity content before treatment to an impurity content after treatment:
a.Uranium from plutonium – at the level of 10 7.
b.Uranium from fission products – at the level of 107.
c.Plutonium from uranium – at the level of 10 6.
d.Plutonium from fission products – at the kevel o f 108.
Such a fine purification gave a foundation to call this approach as the “clean fuel – dirty waste” concept. The concept is very attractive for NFC closure but it can produce some difficulties for nuclear nonproliferation regime. Therefore, when some advanced SNF reprocessing technologies have been developed, which deliberately left some remarkable quantity of radioactive fission products in the recycled fuel to enhance nuclear non-proliferation regime, a new approach arose, namely the “dirty fuel – clean waste” concept.
Nearly 7000 t SNF are discharged annually from the world NPP. Capabilities of the existing and under construction facilities for SNF reprocessing are shown in Table 1.6. As is seen, the existing SNF reprocessing plants are not able to deal with full annual SNF amount discharged from all NPP in operation throughout the world.
SNF reprocessing plant is a rather expensive enterprise. Approximate specific cost of SNF reprocessing is evaluated as 500 US dollars/kg SNF.
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