Performance Assessment of Spent Fuel Assemblies Removed from Shutdown RBMK-1000 Power Units for Reburning in Operating Power Units
B. Kanashov, I. Kuzmin, A. Kostyuchenko, S. Perepelkin, V. Chesanov
X International Nuclear Forum “Safety of Nuclear Technologies: Transport of Radioactive Materials – ATOMTRANS-2015”, St. Petersburg, Russian Federation, 5-9 October 2015
The scheduled operating life-time of RBMK-1000 reactor units, that is 30 years, has already expired. Comprehensive analysis of the reactor conditions allowed to extend the operating life-time of RBMK units up to 45 years . The additional time to operate each reactor will depend on its peculiarities, however, we can predict the NPP will shut down its reactors in the order of their commissioning.
A peculiarity of the RBMK reactors is non-stop core refueling. On the average, one burnt-up fuel assembly (FA) is replaced for a fresh one daily. Sometimes the irradiated assemblies that have not reached the maximum design burnup are reinserted into the core, which already contains the assemblies of different burnups. FA distribution in terms of burnup depends on the duration of the previous period. If the reactor was loaded with the same type fuel with approximately the same speed, the FA burnup distribution in the core does not change. In this case, the refueling is steady-state. Over the period of RBMK reactors operation, the enrichment of fresh fuel has experienced a staged growth. Nowadays, RBMK-1000 units are practically converted to uranium-erbium fuel enriched to 2.8 % in U-235 with erbium content 0.6 %. In particular, the share of assemblies containing the fuel enriched to 2.8 % is 96 % as of the beginning of 2014 .
A shut-down RBMK-1000 reactor will contain many FAs with the burnup far under the design one. This fuel can be used in other reactors. The use of the fuel from a shutdown reactor unit minimizes fresh fuel requirements. Kurchatov Institute specialists studied the feasibility of reburning the fuel from Unit 1 of the Leningrad NPP in other power units of the plant. It was shown that the most cost-saving solution is to reburn fuel assemblies from Unit 1 in Unit 3 and assemblies from Unit 2 - in Unit 4. This does not preclude reburning of Unit 1 fuel in Unit 4 and Unit 2 fuel in Unit 3, i.e. reburning in the Units of the second stage of construction.
In early 2000s, the Ignalina NPP accumulated positive experience in using at Unit 2 the FAs discharged from shut-down reactor of power Unit 1. Positive experience of the Ignalina NPP, as well as Russian experience in FA reburning in the same power unit hold out a hope of safe and cost-effective reburning technology for the Russian NPPs with RBMK-1000 reactors.
1. Operating peculiarities of fuel assemblies at RBMK-1000 power units
The number of FAs with expired operating lifetime unloaded from each RBMK-1000 unit averages 389 spent fuel assemblies (SFA) per year. The assigned lifetime of RBMK-1000 fuel assemblies enriched to 2.8 %wt. (0.6 % of erbium) is 3380 MW day/assembly or 30 MW day/kgU. These parameters make up 4000 MW day/assembly and 35.7 MW day/kgU, respectively, for fuel assemblies with rod height profiling of fuel enrichment. The assigned lifetime for all types of FAs is 10 years provided maximum 8 years of full power operation.
The period of full power operation (PFPO) of a fuel assembly is determined as:
PFPO = P1–P2–P3,
where: P1 is calendar time from the first loading till final unloading of fuel assembly (assigned lifetime), years;
P2 is total time of fuel assembly cooling in the cooling pool, years;
P3 is total time of fuel assembly in a fuel channel of a shut-down reactor, years.
Fig. 1 А shows the structure of the assigned lifetime of RBMK-1000 FAs. The figure does not show the time spent by a fresh FA in the storehouse and on the reactor hall balcony before being loaded into the reactor core. This time is assumed to be less than 0.3 year.
In case of reburning the FAs of a shut-down reactor in the operating one, the structure of the assigned lifetime of RBMK-1000 FA changes (Fig. 1 B). In this case, beside the above-mentioned values, the assigned lifetime includes:
• PFPO -1 is the period of full power operation of the FA in the reactor to be shut down;
• P4 is the time of mandatory cooldown of fuel assembly in the core of the shut-down reactor (minimum 1 year);
• P5 is the time of the FA stay in the core of a shut-down unit during unloading of all FAs (0-4 years).
• PFPO -2 is the period of power operation of the FA in the operating reactor.
The total time of technological storage (P4+P5) of the FAs to be reburnt shall not exceed 5 years: during this period all SNF stored in the core and in the storage pool shall be completely removed from the shut-down power unit in compliance with the power unit decommissioning program. Summing up all periods and assuming that the last unloaded FA of the shut-down reactor shall work in the operating power unit till the end of its service life, we obtain the assigned lifetime of RBMK-1000 FA to be 15 years. Anyway, if an average time of the fuel assembly in the shut-down reactor is 2.5 years, the lifetime of RBMK-1000 fuel assembly shall be assigned minimum 13 years.
2. Estimates of the SFAs fit for reburning
Let us determine the number of fuel assemblies in the Leningrad NPP fit for reburning. Now the reactors of Units 1-4 are mainly loaded with the fuel enriched to 2.8 % in U-235. In 2014, the number of the FAs in Units 1-4 was 1597, 1449, 1591 and 1609 respectively , that sums to 6246 FAs. Suppose that this very number of FAs will be in the reactors of each power unit at the moment of their shutdown. Taking into account sorting-out for several reasons, the share of FAs fit for reburning in other power unit will be considered as 50 %, i.e. approximately 2320 FAs (the Unit 4 FAs cannot be reburnt). Taking into account that each under-burnt fuel assembly can operate in the core of a working reactor maximum for 4 years (1100 eff. days) in the full power mode, the total service life will be about 2550 thousand FA•eff.days.
Reburning of Unit 1 FAs can be started in the beginning of 2020 at best. 798 under-burnt Unit 1 FAs can be used:
in Unit 2 - for 1 year (if Unit 2 is shut down at the end of 2020),
in Unit 3 - for 5 year (if Unit 3 is shut down at the end of 2024),
in Unit 4 - for 7 year (if Unit 4 is shut down at the end of 2026).
The total is 13 reactor-years (or 3575 eff. days).
The fuel assemblies from Unit 2 shut down in 2020 can be reburnt from the beginning of 2022. 725 Unit 2 fuel assemblies can be used:
in Unit 3 - for 3 year (if Unit 3 is shut down at the end of 2024),
in Unit 4 - for 5 year (if Unit 4 is shut down at the end of 2026).
The total is 7 reactor-years (or 1925 eff. days).
795 FAs from Unit 3 shut down in 2024 can be used in Unit 4 since 2026 during one year (if Unit 4 is shut down in the end of 2026). The total is 1 reactor-year (or 275 eff. days).
Potential reburning capacity of Units 2, 3 and 4 is, thus, 3037.1 thousand FA•eff.days. When considering that the reactors of operating power units will be loaded with an equal number of fresh and under-burnt FAs, the operating units left after Unit 1 shutdown in 2018 will provide only the half of this number for reburning, i.e. 1518.5 thousand FA•eff.days.
Comparison with the total service life of under-burnt FAs (2550 thousand FA•eff.days) shows that the number of under-burnt FAs is 1.7 times more than the reburning capacity of the operating power units.
The obtained result indicates that not all under-burnt FAs can be used in the operating power units. On the other hand, the NPP management gets a freedom of choice concerning the use of under-burnt FAs in one or another unit.
3. Requirements for the health of fuel assemblies to be reburnt
Both working FA (Assembly of Type 50) and working FA with gamma chamber (Assembly of Type 49) can be reburnt in the operating power units of the NPP. Taking into account the results of operation and post-irradiation examination, it is suggested not to consider the FAs with zirconium spacer grids as candidate fuel for reburning. Only the FAs with stainless steel spacer grids are allowed for reburning. It is suggested to allow for reburning the FAs with initial enrichment 2.8 % and erbium content 0.6 % (dwg. 865.00.000-16-17) and the FAs with rod height profiling of fuel enrichment (dwg. 871.00.000-01).
As it will be shown below, the FAs having the maximum burnup 20 MWday/kgU (2300 MWday/FA) or having full power operation period about 1100 eff. days (about 4 years) are subject to reburning in the operating NPP units. It is suggested to allow for reburning the FAs after their cooling in the core of the shut-down reactor or storage pool for minimum one year. The FAs having the maximum total service life 13 years, including 8 years of full power operation are allowed for reburning. Post-irradiation examinations of RBMK-1000 FAs having been stored under water in the storage pool for 24 years show that long-term underwater storage is not a restraint the fuel re-use.
The health of the FAs is assessed visually. Particular attention is paid to the state of spacer grids and fuel rod claddings; absence of foreign objects in the inter-rod space is also important. Compliance of a FA with the fabricator's quality reference is established during the visual inspection. A fuel assembly is deemed unfit for reburning, if the following defects are revealed:
• visible damage to the spacer grids (scars, ruptures, dents);
• displacement or absence of the spacer grid;
• visible dents or damage (incl. corrosion) of the fuel rod claddings;
• no fuel rod plug;
• a foreign object in the inter-rod space;
• the gap between the ends of the upper and the lower fuel rod bundles (FRB) is less than 12 mm.
4. The state of RBMK-1000 FA structural materials after operation according to the post-irradiation examination results
4.1. Corrosion resistance of fuel rod claddings
Generally, corrosion of zirconium in water takes place as follows:
Zr + 2H2O -> ZrO2 + 2H2.
At first, the oxide film is oxygen deficient, i.e. it is a ZrO2-x compound, where x = 0.005. This oxide film has high protective properties, good adhesion to the metal sub-layer; it is compact, dark-colored and glossy. The peculiarity of zirconium and zirconium alloys corrosion is that as the oxide film grows, a moment of an abrupt change in its quality and deterioration of its protective properties can happen for a number of reasons. During this period, the oxide film loses its continuity, a system of longitudinal and cross-sectional cracks is generated, the film becomes friable and tends to spall. The oxide layer reaches its stoichiometric composition and takes on marked white or light-gray color. The described phenomena lead to changes in the corrosion kinetics from parabolic to linear. This type of corrosion is observed in the regions between spacer grids. Within the operating time range 650-3100 eff. days, the maximum thickness of the oxide layer in the regions between spacer grids can be described as :
where t is the time of full-power operation, eff. days; t0=650 eff. days; a=0.112 μm/day, b=32 μm.
For 20 MW day/kgU burnup (about 1100 eff. days), the maximum thickness of the oxide layer in the regions between the spacer grids does not exceed 100 μm. The minimum residual thickness of metal is 900 μm (Fig. 2).
Along with continuous (uniform) corrosion, specific types of localized corrosion can be observed on the claddings. So called nodular corrosion appears as small spots (nodules) of white oxide film formed on the cladding surface at some distance from each other, the average diameter of the nodules ranges from fractions of millimeter to several millimeters. In time, the regions of light oxide film can grow larger and merge into a uniform oxidation zone growing into thick laminated oxide film.
The results of post-irradiation experiments [3, 4, 5] showed that localized oxidation and fretting of fuel rod claddings were observed under the spacer grids along with general uniform corrosion. Within the operating time range 527-3100 eff. days, the maximum thickness of the oxide layer in the regions under the spacer grids can be described as :
where: where t is the time of full-power operation, eff. days; t0=527 eff. days; c=10.9 μm/day0.5.
For fuel burnup up to 20 MW day/kgU (about 1100 eff. days), the maximum thickness of the oxide layer in the regions under the spacer grids does not exceed 300 μm (Fig.3). The minimum residual thickness of metal is 700 μm.
During SFAs storage in the neutral water of the storage pool, the rate of general uniform corrosion of the claddings between spacer grids is small, and the oxide film thickness on the fuel rod surfaces depending on the storage period can be described as:
where Вu is fuel burnup, MW day/kgU.
After 5 years of technological storage, additional thickness of oxide film on the claddings of SFAs with burnup 20 MW day/kgU does not exceed 1 μm. This conclusion is confirmed by the results of post-irradiation examination of a SFA that have been stored in the storage pool during 24 years .
However, corrosion of zirconium claddings has other mechanism in the regions with nodular corrosion and under the spacer grids. The regions of the fuel rod surface having traces of nodular corrosion continue to corrode in the storage pool because of absence of protective oxide layer. In such regions, corrosion rate is about 45 times more intensive than in the regions with uniform corrosion, and the oxide film thickness, depending on the storage period, can be described as :
After 5 years of technological storage, additional thickness of oxide film can reach 100 μm in the regions with nodular corrosion and under the spacer grids of SFAs with burnup 20 MW day/kgU. It is unreasonable to regard these SFAs as candidates for reburning in the operating reactors.
4.2. Fretting of fuel rod claddings
The results of post-irradiation examination of more than 20 RBMK-1000 SFAs with varying fuel burnup evidence that fretting of fuel rod cladding under steel spacer grids occurs very rarely, and the thickness of fretting on the large majority of the examined SFAs does not exceed 300 μm. The only case of fretting with the depth of fretting marks of several hundreds of microns, and with the number of these marks up to several tens was observed on the fuel assembly that had worked for 1290 eff. days  till approximately 20 MWday/kgU burnup, was found leaky during in-pile CLT and unloaded from the core ahead of schedule. According to post-irradiation examinations, the leakage was caused by damage by debris of the one fuel rod cladding from the upper bundle.
4.3. Changes in mechanical properties of cladding materials
In-pile irradiation of zirconium alloys during the operation increases their strength and decreases their plasticity. Tests of irradiated cladding samples made of Э110 alloy are given in Fig. 5.
During the operation of RBMK-1000 fuel rods with Э110 alloy cladding till the design burnups:
• ultimate strength increases by 10–20 %;
• yield strength increases approximately by a factor of 1.5;
• total elongation decreases by factor of 3–6;
• uniform elongation decreases by factor of 3–7.
Hydrogen uptake of the cladding material is not observed: hydrogen content in sealed fuel rod cladding ranges within (0.8–1.5)•10-2 %wt. As for changes in cladding material properties during long-term storage, strength and structural characteristics of the cladding materials of RBMK fuel rods after 15-20 years of pool storage are practically the same as those of SFA fuel rods after 2–3 years of storage.
4.4. Changes in mechanical properties of materials of carrier and skeleton rods
In-pile irradiation of Э125 zirconium alloys during the operation increases their strength and decreases their plasticity. Test results of the irradiated samples of Э125 alloy are given in Table 2.
The results from the table indicate that the skeleton rod material has enough strength and plasticity for SFA reburning.
4.5. Changes in mechanical properties of spacer grid materials
Material of the steel spacer grids rim after operation up to 20 MWday/kgU burnup has satisfactory plasticity demonstrated in multiple examinations, as well as by the exterior of the scares formed on the rim during transport operations with SFAs. Post-irradiation examinations did not reveal any significant deformations of spacer grid cells leading to changes in their configuration. Plasticity of steel spacer grid cells material remains high. A limit of elastic displacement, i.e. the maximum increase of the cell diameter without residual plastic deformation when expanding the cell with a conic tool, is used as an indicator of spacer grid cells plasticity. The majority of elastic displacement values were within the range of 0.19–0.37 mm, i.e. the cell of steel spacer grids suffer elastic deformation. Ultimate strength of the spacer grid cell material is within the range of 680–900 MPa. Shear load of the cells in the spot welds is 1200–2000 N which is comparable with the shear load in the initial state.
5. Control of FA geometry and fuel module as a whole
Before inserting a fuel module (FA + suspension) in the core of an operating reactor, the following geometrical parameters of the FA shall be checked:
• the distance between the ends of the upper and the lower fuel bundles shall be not less than 12 mm;
• the maximum diameter of spacer grid rims shall not exceed 79 mm.
Check of the geometry of suspension-to-FA joint is aimed to control two parameters:
• bending of suspension-to-FA joint: the joint bending shall not exceed 0.6 mm for the length of 300 mm. The check shall be performed in two planes set at an angle of 120 degrees;
• alignment of the suspension with the FA: displacement of the suspension is allowed by not more than 0.4 mm relative to the FA axis.
Before inserting the fuel module in the core of an operating reactor, the health of the suspension and the quality of FA-to-suspension joint are checked. The health of the welded joint of Assembly 50 to Suspension 15 is inspected visually examining three spot welds with a 6x magnifying glass. The health of the welded joint of Assembly 49 to Suspension 16 is inspected visually examining the the weld and the heat-affected zone, performing gamma-control of the weld and leak test with helium.
Re-use of the fuel module is allowed if the weld has no defects and meets leak-tightness criteria.
The results of the geometry control are summed up in the form of a formal record. The results of measurements are recorded into the SFA Specifications Sheet.
The FAs are loaded into the core after a positive assessment of the fuel rod leak-tightness and the exterior of the FA recorded in the SFA Specifications Sheet. During the first five days, each of the re-loaded FAs is subject to every-shift individual control with the channel-by-channel CLT system to confirm the FA operability.
Having analyzed the number of the FAs from the shut-down RBMK-1000 units fit for reburning, their health estimated during post-irradiation examinations and having compared this data to the requirements for the FAs, the following can be concluded:
• The analysis of the health of FAs from shut-down reactors having been working to burnup 20 MW day/kgU (1100 eff. days) demonstrates their significant re-burning potential.
• The candidate FA handling is in need of special requirements to be developed and followed during unloading from the shut-down reactor core, selection for reburning, cutting off suspension, transportation, attachment of new suspension and final control of FA at the operating power unit. • At least 2000 FAs from the shut-down Units 1, 2, 3 of the Leningrad NPP can be potential candidates for re-use in Units 2, 3, 4.
• The capabilities of Units 2, 3 and 4, subject to their shutdown schedule for decommissioning allow them to accept only a part of the above-mentioned FA quantity.
• Since the rate of FA acceptance at the operating units is lower than the rate of under-burnt FA supply from the shut-down units, temporary technological storage of these assemblies is also required that can extend the present RBMK-100 FA lifetime (10 years).
• To avoid frequent replacement of FAs because of their expired assigned lifetime, this lifetime shall be extended up to minimum 13 years.
• The total quantity of under-burnt FAs is sufficient for a strict sort-out to guarantee leak-tightness of FAs during the whole period of their re-use.
• Handling operations shall not interfere with the health of FA components (first of all, fuel rods and spacer grids).
1. B. GABARAEV, Yu. CHERKASHOV, A. PETROV, et al., Justification of Extension of RMBK Reactors Lifetime, Atomic Energy, 2006, Volume 100, Edition 4, p.328-335.
2. PETROV A., CHEREPNIN YU., IVANOV A., DMITRIEVA I., Operation of Fuel Assemblies at RBMK-1000 Power Units, Safety, Efficiency and Economics in Atomic Energy Industry (Proc. 9th Int. Conf. Moscow, 2014).
3. KOSTYUCHENKO А, CHESANOV V., ZVIR Е., MARKOV D., MAERSHINA G., Oxidation of RBMK 1000 Fuel Rods Claddings During Standard Operation, Reactor Material Science (Proc. 9th Conf. Dimitrovgrad, 2009), pp. 664–669.
4. SUKHIKH A., KOBYLYANSKY G., MAERSHINA G., NEUGODNIKOV D., State of Fuel Rod Claddings and Spacer Grids of RBMK-1000 Fuel Assemblies with Design and High Fuel Burnup, Reactor Material Science (Proc. 7th Conf. Dimitrovgrad, 2003), pp. 170–179.
5. NEUGODNIKOV D., MARKOV D., KOBYLYANSKY G., MAERSHINA G., Analysis of Post-Irradiation Examination Results of RBMK-1000 SFAs with Uranium and Uranium-Erbium Fuel with High Burnup, Reactor Material Science (Proc. 8th Conf. Dimitrovgrad, 2007).
6. RAZMASHKIN N., KRITSKY V., BEREZINA I., Problems of Long-Term Storage of Spent Nuclear Fuel, Reactor Material Science (Proc. 9th Conf. Dimitrovgrad, 2009), pp. 478–485.
7. PAVLOV S., KRITSKY V., ILYIN P., SHALAGINOVA T., RAZMASHKIN N., Materials Science Issues of Long-Term Wet and Dry Storage of VVER and RBMK SNF, Reactor Material Science (Proc. 9th Conf. Dimitrovgrad, 2009), pp. 455–477.
8. VOROBEY I., MARKOV D., SMIRNOV A., KUZMIN V., LYADOV G., STUPINA L., LESHCHENKO A., BEK E., RYABOV V., Corrosion of RBMK-1000 Fuel Rod Claddings after Standard Operation to Burnup 6.3-19.5 MW day/kgU, Reactor Material Science (Proc. 6th Conf. Dimitrovgrad, 2000), Vol.2, Part 2, pp. 177–185.
9. KOBYLYANSKY G., NOVOSELOV A., Radiation Resistance of Zirconium and Zirconium-Based Alloys, Reactor materials science reference materials, Ed. Tsykanov V., Dimitrovgrad, NIIAR, 1996 – 176 pages.
10. MARKOV D., PAVLOV S., NOVOSELOV A., POLENOK V., ZHITELEV V., ZVIR E., CHESANOV V., KOBYLYANSKY G., New Generation VVER and RBMK Fuel: Post-Irradiation Examination Results, Reliability and Operability Justification, Reactor Material Science (Proc. 9th Conf. Dimitrovgrad, 2009), pp. 24–45.
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