IAEA/USA Interregional course on
Preparations to Ship Spent Nuclear Fuel (1997)
THE PROBLEMS OF TREATMENT
OF IRRADIATED FUEL
AT RUSSIAN RESEARCH REACTORS
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Contact:
Mr. Nikolai Arkhangelsky
Ministry of Atomic Energy of the Russian Federation
Staromonetny, 26
109180 Moscow, Russia
Tel: + 7-095-239-4144
Fax: +7-095-233-3053
E-mail: none
IAEA/USA Interregional Training Course
Technical and Administrative Preparations Required for Shipment of Research Reactor Spent
Fuel to its Country of Origin
13-24 January 1997
Argonne, IL
Lecture L.13.6
THE PROBLEMS OF TREATMENT
OF IRRADIATED FUEL
AT RUSSIAN RESEARCH REACTORS
N.V.Arkhangelsky
Ministry of Atomic Energy of the Russian Federation
TOPICS
- MODERN STATUS WITH RUSSIAN RESEARCH REACTORS
- TEMPORARY STORAGE IN THE POOLS OR IN THE VESSELS
- STORAGE IN THE REPOSITORY ON THE SITE OF THE REACTORS
- TRANSPORTATION OF THE FUEL TO THE REPROCESSING PLANT
- RUSSIAN RERTR PROGRAM, MODERN STATUS
MODERN STATUS WITH
RUSSIAN RESEARCH REACTORS
- 18 REACTORS CONTINUE TO OPERATE
- LARGE VARIETY OF TYPES OF FUEL ASSEMBLIES
- MANY TYPES OF EXPERIMENTAL AND EXOTIC FUELS
- MAIN RUSSIAN INSTITUTES AND FACTORIES IN THE FIELD
OF DEVELOPMENT AND FABRICATING OF FUEL ELEMENTS AND ASSEMBLIES FOR RESEARCH
REACTORS:
- The designer of fuel elements is All-Russian Research
Institute for Inorganic Materials;
- The designer of fuel assemblies is Research and Design
Institute for Power Engineering;
- The main fabricators of
fuel elements and assemblies are the Novosibirsk Chemical Concentrates
Plant and the Machine-Building Plant in Elektrostal;
- The reprocessing plant is "Mayak".
- The designer of fuel elements is All-Russian Research
Institute for Inorganic Materials;
RUSSIAN RESEARCH REACTORS
Facility Name |
Date of Criticality |
Reactor |
Power, |
|
Age, |
Enrichment of uranium, % |
Fuel Type | |
1. |
F-1 |
1946 |
graphite |
0.024 |
Oper. |
50 |
Natural |
U metal |
2. |
IR-8 |
1957 |
pool |
8.0 |
Oper. |
39 |
90 |
UO2 + Al |
3. |
MR |
1963 |
channel |
40.0 |
Shutdown |
- |
90 |
UO2 + Al |
4. |
IIN-3M |
1972 |
homog (l)pulse |
1*10(4) |
Oper. |
24 |
90 |
Solution |
5. |
ARGUS |
1981 |
homog(l) |
0.05 |
Oper. |
15 |
90 |
Solution |
6. |
IR-50 |
1961 |
pool |
0.05 |
Reconstr. |
35 |
10 |
UO2 + Mg |
7. |
IRT - MIFI |
1967 |
pool |
2.5 |
Oper. |
29 |
90 |
UO2 + Al |
8. |
AM |
1954 |
graphite |
10.0 |
Oper. |
42 |
4.4 & 10 |
UO2 + Mg |
9. |
BR-10 |
1958 |
fast |
8.0 |
Oper. |
38 |
- |
UN |
10. |
WWR-TS |
1964 |
pool |
15.0 |
Oper. |
32 |
36 |
UO2 + Al |
11. |
SM-3 |
1961 |
tank |
100.0 |
Oper. |
35 |
90 |
UO2 + Cu |
12. |
MIR-M1 |
1966 |
channel |
100.0 |
Oper. |
30 |
90 |
UO2 + Al |
13. |
BOR-60 |
1969 |
fast |
60.0 |
Oper. |
27 |
- |
UO2+ PuO2 |
14. |
RBT-6 |
1975 |
pool |
6.0 |
Oper. |
21 |
63 |
UO2 + Cu |
15. |
RBT-10/1 |
1983 |
pool |
10.0 |
Temp.shtd |
13 |
63 |
UO2 + Cu |
16. |
RBT-10/2 |
1984 |
pool |
10.0 |
Oper. |
12 |
63 |
UO2 + Cu |
17. |
IVV-2M |
1966 |
pool |
15.0 |
Oper. |
30 |
90 |
UO2 + Al |
18. |
WWR-M |
1959 |
pool |
18.0 |
Oper. |
37 |
90 |
UO2 + Al |
19. |
IRT-T |
1967 |
pool |
6.0 |
Oper. |
29 |
90 |
UO2 + Al |
20. |
RG-1M |
1970 |
pool |
0.1 |
Oper. |
26 |
10 |
UO2 + Mg |
21. |
IBR-2 |
1977 |
pulse |
4.0 (av.) |
Oper. |
19 |
- |
Pu |
22. |
PIX |
- |
tank |
100.0 |
Constr. |
90 |
90 |
UO2 + Cu |
TEMPORARY STORAGE
IN THE POOLS OR IN THE VESSELS
- CAPACITY OF THE STORAGES CORRESPONDS TO 8-10 YEARS
OF OPERATION OF THE REACTOR AND IS ENOUGH FOR PROVIDING OF THE POSSIBILITY
OF THE UNLOAD OF THE FULL CORE TO THE STORAGE IN THE CASE OF ACCIDENT
- PREVENTION OF INADVERTENT CRITICALITY
- PROVIDE AN ADEQUATE COOLING OF FUEL ELEMENTS
- AVOID CORROSION OF THE CLADDING OF THE FUEL ELEMENTS
AND ASSEMBLIES
- STORAGE OF DEFECTIVE FUEL ELEMENTS AND ASSEMBLIES
IN SPECIAL BOXES
- RECORDING AN INFORMATION IN A SPECIAL LOGS
STORAGE IN THE REPOSITORY
ON THE SITE OF THE REACTORS
- DIVIDING ALL REPOSITORIES INTO 3 CLASSES DEPENDING
ON THE POSSIBLITY OF THE REPOSITORY OVERFLOWING IN THE CASE OF SOME ACCIDENTS
- PREVENTION OF INADVERTENT CRITICALITY
- PROVIDE AN ADEQUATE COOLING OF FUEL ELEMENTS
- AVOID CORROSION OF THE CLADDING OF THE FUEL ELEMENTS;
DRY REPOSITORIES ARE PREFERABLES
- STORAGE OF EXPERIMENTAL FUEL ELEMENTS AND ASSEMBLIES IS DIFFICULT PROBLEM
TRANSPORTATION OF THE FUEL
TO THE REPROCESING PLANT
- USING THE TRANSPORT CASK DESIGNED SPECIALLY FOR TRANSPORTATION
OF SPENT FUEL ASSEMBLIES FROM RESEARCH REACTORS
- DECREASING OF DECAY HEATING BEFORE TRANSPORTATION
DURING 3 - 6 YEARS
- PREPARATION OF THE NECESSARY DOCUMENTS
- TRANSPORTATION ONLY UNFAILED FUEL
ASSEMBLIES IN HERMETICAL TRANSPORT CASK
- IT IS POSSIBLE TO TRANSPORT ONLY STANDAR FUEL ASSEMBLIES
INCLUDING IN SPECIAL STANDARD
- TRANSPORTATION OF SPENT FUEL ASSEMBLIESIN ACCORDING WITH SPECIAL PLAN
Main Technical Characteristics of the Cask-19
Characteristic |
Value |
Body wall thickness, mm |
230 |
Bottom thickness, mm |
220 |
Cap thickness, mm |
220 |
Inner cavity diameter, mm |
220 |
Outer diameter, mm |
680 |
Diameter through |
910 |
Inner cavity height, mm |
1430 |
Cask height, mm |
2170 |
Loaded cask mass, kg, not more |
4770 |
Empty cask mass, kg, not more |
4700 |
CASK - 19
- SHIPMENT OF SPENT FUEL ASSEMBLIES FROM PRACTICALLY
ALL RUSSIAN RESEARCH REACTORS
- OPERATION CYCLE OF CASK IS ABOUT 30 DAYS
- EACH CASK CONTAIN 4 FUEL ASSEMBLIES PLACED IN A BASKET
- COOLING TIME IS NOT LESS THAN 3 YEARS
- CASK-19 SATISFIES ALL THE REQUIREMENTS OF IAEA RULES
- TOTAL DECAY HEAT IN ASSEMBLIES IN CASK UP TO 360 W
- THREE MODIFICATIONS OF INNER BASKET FOR DIFFERENT FUEL ASSEMBLIES
RUSSIAN RERTR PROGRAM,
MODERN STATUS
- START OF THE PROGRAM AT THE END
OF SEVENTIETH
- DEVELOPOMENT OF DIFFERENT FUEL COMPOSITIONS
- URANIUM OXIDE IN ALUMINUM MATRIX
- URANIUM SILICIDE IN ALUMINUM MATRIX
- FIRST STAGE OF THE PROGRAM IS OVER - THE MODERN FUEL
COMPOSITION PROVIDE THE DECREASING OF URANIUM FROM 90 % TO 36 %
- ACCORDING TO THE SECOND STEP OF THE PROGRAM THE ENRICHMENT
SHALL BE DECREASE TO LESS THAN 20%
- NOW THE REACTOR EXPERIMENTS ON SECOND STAGE OF THE PROGRAM ARE IN PROGRESS
THE PROBLEMS OF TREATMENT OF
IRRADIATED FUEL
AT RUSSIAN RESEARCH REACTORS
N.V.Arkhangelsky
Ministry of Atomic Energy of the Russian Federation
Staromonetny, 26
109180 Moscow, Russia
Tel. + 7 095 239 41 44
Fax. +7 095 233 30 53
Presented at the
IAEA/USA Interregional Training Course
on the Technical and Administrative Preparations
Required for Shipment of Research Reactor
Spent Fuel to Its Country of Origin
Argonne National Laboratory
13-24 January 1997
ABSTRACT
This report describes the problems of the storage and transportation
of the spent fuel from Russian research reactors.
Many research reactors continue to operate at Russia at present time.
They use many different types of fuel elements and assemblies.
This report discusses three stages of the storage and transportation
of the spent fuel:
the temporary storage in the pool or in the vessel;
the storage in the repository on the territory of the institutes;
the transportation of the fuel to the reprocessing plant.
The future plans provide the solution of the problem of transportation and reprocessing of all types of fuel assemblies which are used in Russian research reactors and experimental facilities. Also the Russian Reduced Enrichment Research Reactors Program that was started late in 70th continuing now. The main results of this work would be increase the density of the fuel meat in the composition on the basis of uranium dioxide and the change of the fuel composition to uranium silicide in aluminum matrix or another.
1. Introduction
From 1946 to the present a large number of research reactors have been
constructed in Russia. Twenty of them continue to operate at the present
time. Moreover there are a certain number of prototype reactors, pulse type
reactors, reactors for special purposes and zero power reactors (or critical
assemblies).
The total power of all operating Russian research reactors is about 400
MW. The value of cumulative reactor years for all of them is more than 600
(see Table 1). During their operational lives the Russian research reactors
accumulated and continue to accumulate a large number of irradiated fuel
elements and assemblies of different types, including the experimental fuel
elements and assemblies.
Naturally, these circumstances required the development of a system of
treatment of irradiated fuel that could guarantee the nuclear and radiation
safety of the storage and transportation of this fuel. Such system has been
created and is continually updated.
The important peculiarity of Russian research reactors is the large variety of types of fuel assemblies that are used in different reactors. At the present time, more than ten types of fuel assemblies are in use. The fuel elements in these assemblies are of tube and rod types, the enrichment of uranium is 10, 36 and 90%, the height of the active part is from 35cm to about 2m (see some examples of Russian fuel assemblies on Figures 1 to 4). Different types of fuel materials are in current use, such as uranium-aluminum alloy, a dispersion of uranium oxide in aluminum matrix, etc.
Table 1
RUSSIAN RESEARCH REACTORS
Facility Name |
Date of Criticality |
Reactor |
Power, |
Status |
Age |
Enrichment of uranium, % |
Fuel Type | |
1. |
F-1 |
1946 |
graphite |
0.02 |
Oper |
50 |
Natural |
U metal |
2. |
IR-8 |
1957 |
pool |
8.0 |
Oper. |
39 |
90 |
UO2 + Al |
3. |
MR |
1963 |
channel |
40.0 |
Shut down |
- |
90 |
UO2 + Al |
4. |
IIN-3M |
1972 |
homog (l pulse |
1*10(4) (in pulse) |
Oper. |
24 |
90 |
Solution |
5. |
ARGUS |
1981 |
homog(l) |
0.05 |
Oper. |
15 |
90 |
Solution |
6. |
IR-50 |
1961 |
pool |
0.05 |
Reconstr. |
35 |
10 |
UO2 + Mg |
7. |
IRT - MIFI |
1967 |
pool |
2.5 |
Oper. |
29 |
90 |
UO2 + Al |
8. |
AM |
1954 |
graphite |
10.0 |
Oper. |
42 |
4.4 & 10 |
UO2 + Mg |
9. |
BR-10 |
1958 |
fast |
8.0 |
Oper. |
38 |
- |
UN |
10. |
WWR-TS |
1964 |
pool |
15.0 |
Oper. |
32 |
36 |
UO2 + Al |
11. |
SM-3 |
1961 |
tank |
100.0 |
Oper. |
35 |
90 |
UO2 + Cu |
12. |
MIR-M1 |
1966 |
channel |
100.0 |
Oper. |
30 |
90 |
UO2 + Al |
13. |
BOR-60 |
1969 |
fast |
60.0 |
Oper. |
27 |
- |
UO2+ PuO2 |
14. |
RBT-6 |
1975 |
pool |
6.0 |
Oper. |
21 |
63 |
UO2 + Cu |
15. |
RBT-10/1 |
1983 |
pool |
10.0 |
Temp.shtd |
13 |
63 |
UO2 + Cu |
16. |
RBT-10/2 |
1984 |
pool |
10.0 |
Oper. |
12 |
63 |
UO2 + Cu |
17. |
IVV-2M |
1966 |
pool |
15.0 |
Oper. |
30 |
90 |
UO2 + Al |
18. |
WWR-M |
1959 |
pool |
18.0 |
Oper. |
3 |
90 |
UO2 + Al |
19. |
IRT-T |
1967 |
pool |
6.0 |
Oper. |
29 |
90 |
UO2 + Al |
20. |
RG-1M |
1970 |
pool |
0.1 |
Oper. |
26 |
10 |
UO2 + Mg |
21. |
IBR-2 |
1977 |
pulse |
4.0 (av.) |
Oper. |
19 |
- |
Pu |
22. |
PIX |
- |
tank |
100.0 |
Constr. |
90 |
90 |
UO2 + Cu |
In the former Soviet Union it was developed three generations of fuel
elements and assemblies for research reactors on the basis of using of various
aluminum materials (alloys and oxides).
Table 2
Generations of fuel elements in Russian research reactors
Generations |
Years |
Enrichment, % |
Concentration of U-235, g/l |
Thickness of fuel elements, mm |
Specific heat transfer surface, m2/l |
First |
1954-1970 |
10-36 |
50 |
10 |
0.098 |
Second |
1963-1985 |
36-90 |
60 |
3.2-2.0 |
0.2-0.362 |
Third |
1972-till now |
90 |
68-130 |
1.4-1.25 |
0.525-0.66 |
The designer of fuel elements for Russian and Russian supplied research
reactors is All-Russian Research Institute for Inorganic Materials, the
designer of fuel assemblies is Research and Design Institute for Power Engineering.
The main fabricators of fuel elements and assemblies for Russian and
Russian supplied research reactors are the Novosibirsk Chemical Concentrates
Plant and the Machine-Building Plant in Elektrostal. The Machine-Building
Plant in Elektrostal fabricates the fuel elements and assemblies only for
SM-3 and AM reactors as the Novosibirsk Chemical Concentrates Plant fabricates
fuel elements and assemblies on the base of the aluminum for all remaining
research reactors.
The only reprocessing plant in Russia is "Mayak" near Chelyabinsk in Ural region reprocess all spent fuel elements from Russian research reactors.
Moreover many types of experimental fuel elements are tested in experimental
loops and rigs. It is clear that this situation means that the safety analyses
of storage and transportation of spent fuel elements requires a great deal
of effort.
2. General
At first, it is important to note that the requirements of safety during
the storage and transportation of spent fuel assemblies are described in
many special Russian documents on the safety of research reactors.
In Russia there is a system of the management and storage of spent nuclear
fuel at research reactors. The documents on safety of research reactors
describes of the main features of this system. The top level document is:
"General Requirements for Providing Safety at Research Reactors."[1]
This document includes the main safety principles for the siting, design,
commissioning, operation, modification and decommissioning of research reactors.
It corresponds to such Agency documents as "Safety Standards on Design
and Operation" (35-S1 & 35-S2).
The documents of the next level that describe the definite requirements
for the safety systems of reactor are very numerous. They consider all of
the general questions of the safety during the storage and transportation
of spent fuel elements.
The following main safety requirements shall be provided in all stages
of the movement of spent fuel from the core to the reprocessing plant:
provision of sufficient subcriticality in the storage of the spent fuel elements and assemblies for nuclear safety;
adequate heat removal from the spent fuel elements for the prevention of damage of the cladding of the fuel elements;
provision of an adequate coolant chemistry for the minimization of cladding corrosion;
physical protection of the nuclear fuel;
exact registration of the quantity of nuclear materials at all stages
of the storage and transportation.
3. Temporary Storage of Spent Fuel at the Site
of the Reactor
After unloading the fuel assemblies from the active core they shall be
transported to the temporary storage. All research reactors even the very
low power reactors have a facility for temporary storage of spent fuel.
Rates and order of transportation and storage of the spent fuel elements
shall be specified in the reactor design. A special document that discusses
the problems of safety shall, be included in the design of the reactor.
This document is entitled: "The Technical Basis for the Safety of the
Reactor" and its contents are analogous to those of the "Safety
Analyses Report." The Technical Basis for the Safety of the Reactor
such as its design shall be approved by the regulatory body.
The assemblies of spent fuel elements must be stored in special storage
racks. These storage racks are located in the pool or in the vessel of reactor.
Transfer of the elements from the active core to the storage racks is accomplished
by moving the fuel assemblies through the water that provides adequate shielding.
When the storage racks become full the spent fuel assemblies must be loaded
into the temporary "wet" storage. Usually the capacity of the
temporary storage is relatively big and corresponds to 8-10 years of operation
of the reactor. In all cases the capacity of temporary storage must be enough
for providing of the possibility of the unload of the full core to the storage
in the case of the accident.
The storage of spent fuel assemblies shall be such as to prevent inadvertent
criticality and to provide an adequate cooling of fuel elements. The most
important safety parameter during the storage of the spent fuel assemblies
is a step between fuel assemblies. Its value shall be calculated by the
operational organization. According to the regulations [2] the subcriticality
of spent fuel storage shall not be less than 0.05 in all accident situations
such as mechanical damage of the facilities, loss of coolant in the storage,
special internal and external events etc. These calculations shall be verified
by the special division of the Institute of Physics and Power Engineering
at Obninsk and after that be approved by the regulatory body.
To avoid corrosion of the cladding of the fuel elements the water chemistry
in the storage pool shall be the same as in the primary circuit of the reactor.
For example, for fuel elements with aluminum cladding the value of pH shall
be from 5.5 to 6.5 and the specific conductivity shall be less than 1.5
microS/cm. Clearly, these parameters are also defined in special standards.
Conservative assumptions are used to provide an adequate safety margin.
In particular, the calculations of criticality are carried out assuming
that all fuel elements in storage are fresh, but in the calculations of
the radiological consequences it is assumed that all fuel elements have
the maximum possible burnup.
A special problem is the storage of defective fuel elements and assemblies,
since there may be a possibility that fission products are released into
the pool. These fuel assemblies must be install in special boxes to isolation
them from the pool. However, it is emphasized that occurrences of failured
fuel elements are very rare and at the majority of Russian research reactors
such failures have not happened more than one or twice during the whole
history of the reactor.
Information about all transport operations with spent fuel assemblies
shall be recorded in a special log. These include the necessary information
about fresh fuel assembly (in particular, the date of the construction of
the fuel assembly, the charge of uranium in the fuel assembly, the enrichment
of the uranium), the times of loading and unloading of the fuel assemblies
from the core, the burnup of the fuel and the position of the fuel assembly
in the storage. The order of the treatment of the fuel is described in the
instructions for the safe transportation and storage of fresh and spent
fuel, a document that is available at every research reactor.
4. Storage of the Spent Fuel in the Repositories
Once the decay heating has reduced to a level when the storage of the
spent fuel assemblies is possible without cooling by pool water they can
be transported to the repositories. Such a repository is always available
at all institutes that have several research reactors and in several cases
at the institutes having one research reactor.
The questions of nuclear safety during the transportation of the spent
fuel assemblies to the repositories and storage of spent fuel assemblies
in them is defined in the special document entitled: "Nuclear Safety
Regulations for Transportation and Storage of Dangerous Fissionable Materials"[3].
According to these regulations, all repositories shall be divided into 3
classes depending on the possibility of the repository overflowing in the
case of the flooding, failures of equipment in the neighbouring rooms and
personnel errors. In the repositories of the first class overflowing is
impossible even in the case of the flooding. It is obvious that these repositories
are dry. In the repositories of the second class the penetration of subterranean
waters in the repository is impossible in all cases. These repositories
are also dry. The new repositories shall be designed only as the repositories
of the first or second classes.
As we see from the name of this document it determines mainly the questions
of nuclear safety. According to it, the subcriticality of spent fuel in
repositories shall not be less than 0.05 for full repository and in all
accident situations. The possible initial events of the accidents are earthquakes,
flooding, fire, loss of electrical power supply, failure of transport cask,
etc. The calculations of criticality must be carried out by the operational
organization, then they shall be verified by special division of the Institute
of Physics and Power Engineering at Obninsk and after that shall be approved
by the regulatory body. This procedure is the same as for temporary storage
of spent fuel assemblies.
The problem of corrosion of the cladding of fuel elements can arise during
the storage of spent fuel assemblies in "wet" repositories but
it is important to emphasize that the process of corrosion proceeds very
slowly. However, the long-term storage of spent fuel assemblies in dry repositories
shall be preferable.
The special problem is the storage of the spent fuel assemblies from
experimental loops and rigs. It's necessary to say that the many of them
were designed on unique technology that did not receive the continuation.
In several cases we must to say about a few fuel elements.
5. Transportation of Spent Fuel to the Reprocessing Plant
It is necessary that before transportation of the spent fuel assemblies from the institutes to the reprocessing plant the level of decay heating become enough low.
The value of necessary time for the decreasing of decay heating varies from one fuel
assembly to another and may reach six years for fuel assemblies with
very high specific power. The special standards determine these values of
the times for every type of fuel assembly.
After the decay heating in fuel elements has reduced to a low level they
can be transported to the reprocessing plant. The questions of nuclear safety
during the transportation of the fuel assemblies are defined in special
document: "Principal Safety and Physical Protection Regulations for
the Transportation of Nuclear Materials" [4].
The transportation of spent fuel elements carry out in special dry containers -cask-19 [5]. The cask-19 has been developed specially for research reactor spent fuel assemblies transport. It is a thick-walled vessel tightly covered with a heavy cap.
The cask allows the shipment of spent fuel assemblies of different research
reactors with various fuel types of different cross-sectional shapes and
with a total decay heat of up to 360 W. The cask design assures nuclear
safety, biological shielding, heat removal from fuel assemblies, hermetic
sealing and strength in normal and emergency conditions of transportation.
Main technical characteristics of the cask are given in Table 3.
The cask-19 design satisfies all the requirements of IAEA rules on B
(V) - type package safety. The cask-19 leak tightness is provided by the
use of packing in the sealing units, made of rubber based on silicone as
fillers and allowing short-term operation at 250 C in emergency conditions.
Table 3. Main technical characteristics of the cask-19.
Characteristic |
Value |
Body wall thickness, mm |
230 |
Bottom thickness, mm |
220 |
Cap thickness, mm |
220 |
Inner cavity diameter, mm |
220 |
Outer diameter, mm |
680 |
Diameter through trunnions, mm |
910 |
Inner cavity height, mm |
1430 |
Cask height, mm |
2170 |
Loaded cask mass, kg, not more |
4770 |
Empty cask mass, kg, not more |
4700 |
The cask-19 allows the shipment of spent fuel assemblies from practically
all Russian research reactors excluding AM, BOR-60. There are another transport
cask for transportation of the spent fuel elements from these reactors.
Standard loading of fuel assemblies into the cask-19 should be performed
in the cooling pool under protective water layer.
Transport of the cask-19 loaded with fuel assemblies is done in a railway
cask-car in vertical position. The number of cask-19 loaded into a car depends
on the spent fuel cooling time and spent fuel assemblies characteristics.
The cask operation regime is periodic. One operation cycle is about 30
days. The operation cycle includes: fuel charge, transport, discharge, preparation
to a new cycle.
In the cask inner cavity a basket is situated. Because of the design
differences of fuel assemblies shipped, the basket is made in three modifications
differing by the basket inner cavity alone. As a rule, each cask may contain
4 fuel assemblies placed in a basket. The cooling time of spent fuel to
be transported depends on the burnup, initial enrichment and fuel exposure
time and, as a rule, is not less than 3 years.
The cask-19 may be used in the ambient air temperature from - 50 C to
+ 38 C with relative humidity 90% at + 25 C.
The average service life of a cask is not less than 20 years.
It is impossible to transport spent fuel assemblies until the operational
organization have prepared the necessary documents. These documents shall
consider the questions of nuclear and radiation safety, they shall include
the calculations of criticality and adequate cooling in normal conditions
and in all accident situations. It's possible to transport only the unfailed
fuel assemblies or failed fuel assemblies in hermetical transport cask.
To receive permission for the transportation of spent fuel assemblies,
an operating organization must prepare the necessary documents that must
then be approved by the regulatory body, representatives of security service
and local authorities of those territories over which the transportation
is intended.
The procedure to receive permission is not very easy. Moreover the permission
can be received only if the reprocessing plant has adequate technology for
reprocessing of the spent fuel in question. Taking into account the large
number of fuel compositions that are used in Russian research reactors and
their experimental facilities this problem is very complicated and for this
reason many types of fuel elements stay in the repositories for many years.
There is a special standard for transportation of spent fuel assemblies from research reactors to the reprocessing plant. It includes all main fuel assemblies such as IRT-M, WWR-M, CM-2, MR fuel assemblies with EK-10 fuel elements. There is an adequate technology for reprocessing of these spent fuel assemblies and during several years this technology is use. Concerning the experimental fuel assemblies it is necessary to emphasized that the many of them were designed on exotic technology and this technology did not receive the continuation.
Special standards describe the requirements for the vehicle and the transport
container. New transport containers shall be tested by the design organization
and after the results of the test shall be approved by the regulatory body.
In the former Soviet Union there was a special plan of transportation
of the standard spent fuel assemblies from research reactors to "Majak".
Now the quantity of the spent fuel assemblies at the sites of research reactors
are very big but the financial possibilities of the institutes can not allow
to transport of the spent fuel assemblies to reprocessing plant.
It is possible to construct a new storage at the site of the reactor
but it is also a very expensive decision. The one real way of temporary
solving of the problem of spent fuel is increasing of the capacity of existing
repositories by means of decreasing of the step of placing of spent fuel
assemblies .
6. Future plans
The most important difficulties with the treatment of irradiated fuel
from Russian research reactors are the large variety of the fuel elements
and assemblies used in the research reactors and their experimental facilities.
For this reason in the near future we intend to finish the development of
the safety standards and include in the final document standards for all
of the types of the fuel elements and assemblies used in the past and at
present. On the other hand, we want to use in the future not more than one
or two fuel compositions in the fuel elements of research reactors. This
will make it easier to solve the problems of reprocessing of spent fuel.
But as before the reprocessing of experimental spent fuel assemblies would
be remained the big problem.
The technical problems of storage and transportation of standard spent
fuel assemblies from Russian research reactors to the reprocessing plant
solved and only the bad financial situation prevents to the free of the
storages at the site of research reactors.
We also propose to convert the composition using in the fuel elements
of Russian research reactors for another reason. This is in connection with
the problem of the reduction of enrichment of uranium in research reactors
in accordance with the joint international efforts in nonproliferation policy
[7].
At the present time the main type of the fuel composition in our research
reactors is uranium oxide in aluminum matrix. With this composition it is
possible to reduce the enrichment of uranium from 90% to 36%, but to reduce
the enrichment to less than 20% it is necessary to use a new fuel composition
e.g. uranium silicide in aluminum matrix or another composition.
The development of this composition in Russia is now in progress and we hope in the near future to begin using fuel elements with such a composition. Clearly, this development will create new problems of safety during the storage, transportation and reprocessing of spent fuel elements. In particular the reprocessing of spent fuel assemblies can be very complex problem.
REFERENCES
1. "General Requirements for Providing Safety at Research Reactors.",
OPB IR-94.
2. "Nuclear Safety Regulations for Research Reactors", PBJA-03-75.
3. "Nuclear Safety Regulations for Transportation and Storage of
Dangerous Fissionable Materials", PBJA-06-09-90.
4. "Principal Safety and Physical Protection Regulations for the
Transportation of Nuclear Materials", OPBZ-83.
5. KRITZKIJ V.G., MIKHAILOV G.V., TIKHONOV N.S., MAKARCHUK T.F., IVANOV
V.I. "Research Reactor Fuel Handling" in "Management of Spent
Fuel from Research and Prototype Power Reactors and Resides from Post-Irradiation
Examination of Fuel", IAEA-TECDOC-513, Vienna, 1989, pp.29-37.
6. ARKHANGELSKY N.V. "The Problems of Treatment of Irradiated Fuel
at Russian Research Reactors in "Management and Storage of Spent Nuclear
Fuel at Research and Test Reactors", IAEA-TECDOC-900, Vienna, 1996,
pp. 127-138.
7. ADEN V.G., ARKHANGELSKY N.V., STETSKIY Y.A., YENIN A.A. et al. "The Current State of the Russian Reduced Enrichment Research Reactors Program". The report presented to the 1994 International Meeting on Reduced Enrichment for Research and Test Reactors. Williamsburg, Virginia, USA.