1. A.A. Bochvar Institute of Inorganic Materials (VNIINM)
Development of Dispersion Type Fuel Elements for Floating Nuclear Power Plants (FNPP) and Small size Reactors
A.A. Bochvar Institute of Inorganic Materials (VNIINM)
Moscow, Russia
A. Vatulin, A. Savchenko, G. Kulakov etc
KAERI-2010, February

2. Introduction
When considering the park of low power reactor plants that were under design in our country during the last 15–20 years one can see a very long list of RP of different types in the range from 1 MW to 150 MW.
The trouble of the branch is a large variety of reactor plants (more than 40 types and sizes) from which only two plants were granted licenses by Gostechnadzor: KLT- 40С of the water pressurized type intended for Floating NPP and АТУ-2 of uranium-graphite channel type for the 2nd phase of Bilibino NPP.
The high scientific and technical potentialities of Russia allow it to become a leader in this segment of the world market of electrical energy generation. However, the basic criterion of the participation is the availability of operating power units which allows demonstration of their technical abilities and spheres of application.
Unfortunately, aside from FPU with RP KLT-40C designed by OKBM nothing might be proposed to this market.
That is why, the management of Rosatom took the decision to build the head (pilot) FPU having two RP KLT-40C.

3. KLT-40 Type Reactor
High quality of KLT-40 nuclear system providing enhanced safety is achieved on the basis of creation and operation experience of Soviet nuclear-powered icebreakers similar equipment and systems.
Trouble-free operation of the Soviet nuclear-powered icebreakers reactor plants exceeded 100 reactor/years which proves high quality of design decisions.
In the design of the KLT-40 particular attention was focused on the issues of reliability and safety provision consideration of a variety of design basis and off design basis accidents and external impacts on the RP.

4. KLT-40 Type Reactor Safety Concept
Inherent self-protection – basis of safety
Reactor inherent safety properties are due to:
negative reactivity coefficient of the core
large accumulating capacity;
elimination of large diameter pumps in primary circuit;
insertion of special devices in reactor ducts limiting coolant discharge from the reactor on pipeline ruptures.

5. KLT-40 Type Reactor
Defense in Depth – guarantee against radioactive releases.
A number of segmental barriers are provided in reactor.
Localizing systems prevent radioactive products from spreading beyond the containment during all accidents including maximum diameter piping guillotine rupture.
Accident prevention: no failures – no accident.

6. Developments of Low Enrichment Uranium Fuel Elements for Assembly Core of RP KLT-40С
The KLT-40 reactor cores contain fuel rods based on highly enriched (more than 20 % uranium-235) nuclear fuel. To provide the export potential to NP SP with KLT-40 type RP nuclear fuel has to be developed having not more than 20 % enrichment of uranium which complies with the IAEA requirements for the elimination of a risk of the unsanctioned proliferation of nuclear weapons.
However, the NP SP core has not only to meet the IAEA requirements for the non-proliferation but also it has to provide for the economic characteristics higher than those of the atomic icebreaker core. That is why, based on the RP KLT-40 a novel core KLT-40C was designed.
Thus, the designers faced the specific problem to develop high density fuel and fuel elements on its base for use in the assembly core of the reactor plant KLT-40C of the leading FPU.

7. Comparative Core Characteristics of Cores for Atomic Icebreaker and for RP KLT-40С
Channel Core M
Assembly Core
14-14
1. Power resource, TW·h
2.1
2. Core height, mm
920
1200
3. Specific power density, kW/l
150
119
4. Maximal enrichment of uranium per core, %
HEU
15.7
5. Maximal burnup, g/cm3
0.98
0.74

8. Channel of icebreaker core 14-10-3М Assembly for KLT-40С core
Burnable absorber rod
CPS
Fuel element
Space grid posts
Fuel element
Burnable absorber rod
Channel of icebreaker core М
Assembly for KLT-40С core
Cross-Section of FA

9. To ensure the characteristics of the assembly core a fuel element has to contain not less than 6.05 g U per cm3 in the fuel composition.
A.A. Bochvar Institute (VNIINM) was entrusted to develop such a fuel element.
The intricacy of designing low enrichment nuclear fuel for the assembly core of leading FPU consisted in the fact that according to the approved schedule the first two complete sets of the cores had to be delivered in quarters 2 and 3, 2009.
The requirements for fuel element are most fully met by dispersion fuels of propulsion reactors designed by Bochvar Institute and proved well in the atomic icebreaker cores. The modification was made by application of dioxide uranium fuel (instead of intermetallic) in Al matrix alloy as well as Zr alloy cladding. Therefore, FNPP core became more cost-effective and service life will be increased. Besides novel fuel meets the non-proliferation requirements.

10. Core Microstructure of Fuel Rod

11. Production of UO2 Granules General View of Sintered UO2 Granules
Under the plant conditions the method was introduced used to produce UO2 granules from commercial powders of UO2 having different enrichments.
Specifications for granules of UO2 (ТУ ) have been worked out and are in operation.
The process makes it feasible to regulate with any assurance the sizes of “wet” spheroidized UO2 granules, to account for a shrinkage in the process of sintering and practically fully eliminates rejections due to granule sizes.
General View of Sintered UO2 Granules
The total yield of sound granulated UO2 amounts to %.

12. Series of Out-of-Pile Testing Cermet Fuels
The following out-of-pile tests and examinations were carried out:
microstructural investigations of fuel element cross-sections after fabrication;
thermal cycling fuel elements in temperature ranges from 100 to 350 ºС;
long-term thermal tests of fuel elements at 500 ºС and 550 ºС;
measurements of meat thermal conductivity before and after thermal cycling;
thermal tests of elements upon heating to 800 ºС and 10 min holding (simulated off-design accident).
thermal conductivity measurements of unirradiated fuel composition VS temperature and UO2 volume content of meat as well as measurements of cermet fuel linear thermal expansion coefficient the value of which amounted to (14-15)·10-6 deg-1.

13. Series of Out-of-Pile Testing Cermet Fuels
Main Results
Thermal cycling fuel elements within ºС at the quantity of the cycles up to 200 does not lead to structural changes in fuel elements. A good metallurgic bond is retained of cladding and fuel particles with meat; cracks in fuel element meat are not available. а b
Fuel element microstructure before and after thermal cycling tests:
а – before thermal cycling; b – after 200 thermal cycles

14. Series of Out-of-Pile Testing Cermet Fuels
Main Results
Studies of compatibility of UO2 particles with silumin at 550 °С and holding time up to 1000 h have shown that granules did not essentially interact with silumin.
The simulation of an off-design accident has shown that between UO2 particles and the matrix alloy no diffusion interaction takes place up to 750 °С.

15. Series of Out-of-Pile Testing Cermet Fuels
UO2 content of meat
Thermal conductivity, W/(mC)
Temperature, C
Variation in thermal conductivity of unirradiated fuel composition VS temperature and volume content of UO2

16. Postirradiation examinations revealed the following:
Series of in-Pile Testing and Post-Irradiation Examination of Cermet Fuel Elements
The positive results of the out-of-pile investigations of fuels allowed the start of in-pile experiments with shortened fuel elements in FA “Girlyanda” to study the fuel behaviour under irradiation and to determine its characteristics, namely, swelling, thermal conductivity, thermal expansion and others.
Postirradiation examinations revealed the following:
a good metallurgical bond is retained between cladding and fuel composition (FC);
the maximal increase in the fuel element diameter makes up ~ % at the accumulation of 0.98 g/cm3 ;
the swelling of the cermet FC does not exceed ~ 12 %/(g/cm3) at the fission fragment accumulation up to 0.98 g/cm3.

17. Macrosections of Monolithic Fuels
Control fuel element
N 287/15
Fuel element
N 287/03; 0,62 g/cm3
N 287/19; 0,75 g/cm3
In monolithic fuel elements the volume fraction of UO2 in the meat made up 68 % which corresponds to 6.24 g U per unit of meat volume.

18. Uranium intermetallide а – near cladding-meat boundary; b – in meat
Silumin
Pores in matrix
UO2
Uranium intermetallide а b
Microstructure of FC in monolithic fuel element at burnup of 0.89 g/cm3:
а – near cladding-meat boundary; b – in meat

19. Series of in-Pile Testing and Post-Irradiation Examination of Cermet Fuel Elements
In addition to the standard tests of fuel elements within the “Girlyanda” fuel assembly special-purpose investigations of irradiated fuel elements were carried out including measurements of thermal conductivity of irradiated fuel, tests of irradiated fuel elements with purposely made defects in the reactor MIR special loop as well as thermal tests.

20. Series of in-Pile Testing and Post-Irradiation Examination of Cermet Fuel Elements
The Figures illustrate the appearances of reactor “MIR” tested fuel elements having purposely made defects. а b
The weighing of the cermet fuel elements has revealed that no decreases in the mass took place. The cermet fuel has required corrosion resistance.

21. Series of in-Pile Testing and Post-Irradiation Examination of Cermet Fuel Elements
Thermal test
N of fuel element
Temperature of depressurization С
Conditions of experiment
Fuel element state
287/29
680
Continuous heating
Swelling at the bottom
287/32
750
Swelling in the middle
Fuel element N 278/32, 750 С, appearance after thermal test.
Breakage in the middle

22. Series of in-Pile Testing and Post-Irradiation Examination of Cermet Fuel Elements
Macrostructure in leaky zone, fuel element 278/33
Microstructure in leaky zone
Fuel element
287/33,
boundary with cladding
Fuel element
287/33,
centre

23. Series of in-Pile Testing and Post-Irradiation Examination of Cermet Fuel Elements (Conclusions)
The results of the investigations have demonstrated that the central high uranium content fuel is serviceable to the maximal temperature of 600 С.
Out-of and in-pile examinations show that the cermet fuel has good operating properties and is promising for the assembly core of KLT-40C reactor.
The obtained results were used in 2007 in the preliminary design of the fuel element for the head (pilot) FPU core.

24. Operation Conditions and Requirements for Safety
Fuel Element
Operation Conditions and Requirements for Safety
1. Intended life time – h.
2. Intended service life – 4 years.
3. At the nominal power fuel elements operate under conditions of coolant surface boiling. The maximal temperature at fuel element surface (no account for deposits) – 334 °С.
4. Heat density of fuel element surface in operation at nominal power does not exceed 1.38 МW/m2, LHGR of fuel element does not exceed 295 W/cm.
5. Maximal accumulation of fission fragments in fuel composition – 0.75 g/cm3.
6. Increase in fuel element dimensions without account for tolerances for design technology parameters has not exceeded the following values:
 circumscribed diameter of fuel element – not more than 3 %;
 overall length of fuel element – not more than 1 %.
The fuel elements have to ensure the established power generation of 2.1 ТWh at any combinations of above indicated parameters without leakage.

25. Fuel Element Parameters
Characteristic
Value
1. Outer diameter
6.8
2. Overall length
1275
3. Core length (nominal)
1200
4. Fuel composition
UO2+ silumin
5. U235 Enrichment
15.7 and 13.0
6. Cladding from E-110 alloy, diameter, mm
6.8×5.8

26. Conclusion
1. At VNIINM the design, fabrication process and methods of controlling cermet fuel element were developed for the head FPU core.
2. A series of out-of-pile examinations of the fuel elements were carried out with the determination of their characteristics in unirradiated state.
3. In the loop of the reactor MIR (SSC RF RIAR) under way are successful tests of the designed fuel elements within the irradiation facility “Girlyanda”. Two units have been tested, the tests of three more units are in progress; all fuel elements are leak tight. Post irradiation examinations of fuel elements that reached the burnup of (0.89‑0.98) g/cm3 (129‑153 МW·d/kg U) were carried out; they attested their reliability and serviceability under conditions of their operation in core

27. Conclusion
4. Thermal tests of irradiated fuel elements were carried out; the behaviour of irradiated fuel elements having purposely made defects of their claddings was investigated in the reactor MIR loop. The results of the investigations have demonstrated that the cermet fuel have high irradiation resistance under conditions of off-design accidents and corrosion resistance of leakers.
5. The stage of the production preparedness at OAO MC3 and OAO “ChMP” is such that allows the production of fuel elements for the head FPU core in 2009 in a short period of time and with small backfits.
6. On the basis of the results of the design-process studies, out-of and in-pile examinations in 2007 the preliminary design of fuel element was issued, coordinated and approved.