ORIG. RUSSIAN: OPERATING EXPERIENCE OF THE ATOMIC POWER PLANT ON THE ICEBREAKER 'LENIN'

• Document Type: CREST
• Collection: General CIA Records
• Document Number (FOIA) /ESDN (CREST): CIA-RDP88-00904R000100100015-4
• Release Decision: RIPPUB
• Original Classification: U
• Document Page Count: 19
• Document Creation Date: December 22, 2016
• Document Release Date: August 12, 2009
• Sequence Number: 15
• Publication Date: May 1, 1964
• Content Type: STUDY

 Third United Nations
International Conference
on the Peaceful Uses
of Atomic Energy
Confidential until official release during Conference
OPERATING EXPERIENCE OF THE ATOMIC POWER
PLANT ON THE ICEBREAKER "LENIN"
May 1964
Original: RUSSIAN
I.I. Aphrikantov, N.M. Alordvinov, P.D. Novikov, B.C. Pologikh, A.K. Sledzjuk, N.S. Khlopkin,
N.Y. Tsarev
icebreaker "Lenin" precluded the destruction menace to a number of the ships in the Vilkitsky
case forced winter drifting of transport ships in ice fields is excluded, and the number of
accidents and the possibility of ships being lost under ice pressing conditions are also
brought down.
In this respect the successful, piloting of ships through the Arctic region in 1960 when the
increase of a ship piloting rate under ice conditions and to navigation time extension. In this
the icebreaker and to operate it at full power during prolonged period of time in combination
with more reliable use of the ship under arduous ice conditions.
The use of the powerful icebreaker along the Northern Sea Route contributes to substantial
no small importance that include possibilities to maintain oractically stable displacement of
chemical fuel is impracticable. In addition, NPP application provides operational advantages of
and sailing autonomy, with a hull of limited sizes, that provide sufficient icebreakability and
manoeuvrability. To combine these advantages for an icebreaker with a power plant using
The use of nuclear power plarft (NPP) made it possible to build the icebreaker of great power
In October of 1961 the icebreaker "Lenin" brought to the starting point the personnel and
equipment of the drifting station "North Pole 10".
other icebreakers the "Lenin" has piloted along the Northern Sea Route more than 300 :;hips.
had covered approximately 60,000 miles, about 40,000 of which in ice field. In conjunction with
Since that time she has participated annual arctic navigation. By the end of 1963 the icebreaker
The atomic icebreaker "Lenin" (see Fig.1) joined the Soviet Arctic fleet on December 3,1959.
of the ship. Vibration-and shock-proof equipment of NPP provides reliable operation of the
easily controlled over the whole range of sharp load changes and meets all power requirements
classified in marine practice as the heaviest has shown that the nuclear plant is stable and
The icebreaker "Lenin" operation over a long period of time under such conditions which are
strait is extremely significant.
plant under both ice blows and pitching and rolling.
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Now the icebreaker reactors are running with the second fuet charge. The reactors were
refuelled in spring of 1963. Each of the three reactors operated with the first charge more than
11,000 hours and produced 430,000 - 490,000 thermal megawatt-hours. An average burn-up per
a reactor core was 11,000 - 13,000 MWd/tU and the maximum about 30,000 MWd/tU. During
this period fuel elements operated in the primary circuit water for approximately 30,000 hours.
The icebreaker's reactors have shown stable running at various power levels, including the
maximum of 90 MW. The simultaneous operation of the three reactors with an output of 65 MW
each has provided the icebreaker's full horse-power of 44,000 S.H.P.
GENERAL OPERATIONAL CHARACTERISTIC OF NPP
The atomic power plant is about 3100 tons of gross weight including the biological shield and
is designed to produce 360 tons of steam per hour at a pressure of 28 kg/cm2 and a temperature
of 300 - 310?C. Figure 2 shows the general view of the plant.
The principal scheme of NPP plant is given on figure 3. The primary circuit consists of
three separate sections, e-ch including the following equipment: a reactor, two steam generators,
four main circulating pumps, two emergency circulating pumps, four pressurizers and two ion-
exchange-filters. Each section has two loops - bow and stern ones, that is convenient for both
plant operation and piping system and equipment maintenance.
During basic working conditions of a reactor two loops are running. In this case one main
circulating pump of each loop is in the "on" condition and another is in the "ready-for-action"
condit_i,on.
At about 50-megawatt power level the scheme provides reactor operation with one running
loop when one main and one emergency circulating pumps of the loop are on. The course of NPP
operation has confirmed its designed characteristics that can be seen from the following data
for the icebreaker operation at full power.
Note: fraction numerator refers to a bow loop, fraction denominator refers to a stern loop.
The pri,rary.-circuit pump rate-has proved to be-slightly higher than the designed value. In
accordance-with it water heating in reactor was reduced as it can be well seen from figure-4,
where the calculated (1) and experimental (2) graphs of the reactor inlet and outlet temperatures
against reactor power are-shown.
In 1961 reactor temperature -conditions were-changed to approach them to self-regulating
when a temperature effect is in practice-fully compensated by the Doppler effect.
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Table 1
Designed
values
or 65 MW
Operation data for reactors
No.l_ No.2. No.3
Water flow in primary circuit loops,
435
458
435
t/h r
430
467
453
React
ratur
outl
t te
f
r l
o
s
oy
317
311
312
311
or
e
mpe
p
,
e
o
o
313
Reactor inlet temperature for loops, 0
261
260
261.
260
Steam output for loops,t/hr
43L
42
47
43
43,3
42
42
46
Steam pressure
kg/cn12
32
31.5
31
,
31
30.5
31
Steam temperatv-e, OC
310
308
308
309
308
308
Power value determined from
69.6
75.4
73.1
primary circuit parameters, %
Power value determined from
secondary circuit parameters
As one can see,from figure'4, reactor outlet temperature at all power levels is practically
constant (3) under self-regulating conditions.
During reactor core life the characteristics, corresponding to self-regulating conditions,
slightly change due to temperature and Doppler effect. However it does not result in neces-
sitating control system alteratiron.
The-plant design provides the possibility of steam supply for all users from the-general ship
line: main t'irbine-units, electrical stations, evaporative units, etc. With the-aid of valve
system this general steam line,can be separated into sections with the groups of steam genera-
tors from one,or two reactors which provide steam separately to the-bow and stern echelons of
equipment.. The experience gained has shown that the operation of all the steam generators for
the, general steam line has certain advantages over the-above-mentioned echeloned-operation.
Reactor scram during an echeloned steam. supply results in stopping steam provision to the
users fed from the steam generators of this reactor; while with the general steam line-running it
turns out possible to maintain the required pressure inside,it due~to power increase-of the,other
reactors therefore there is no necessity in full disconnection of the steam users.
Owing to this the number of valve system switches under transient and emergency conditions
of the ~ plant operation goes down and hence ~ the -probability of false operations during the, most
serious periods of plant running is reduced.
The, operational experience - has shown sufficient reliability of the - icebreaker's emergency
measures for providing a regular electric power supply to NPP from two electric stations of
3000 kW each. In case,of voltage-loss by one of the stations two emergency diesel genera.
tors of 100 kW each are automatically started and a 1000 kW-reserve diesel generator is
started by hand.
By the, proposal of the-operational personnel the, designed, two-boarded relay-contact circuit of
power supply for reactor control and safety system has been substituted by a semiconductor
valve circuit to up the plant reliability. _ 3 _
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It should-be-noted-that there-has been no cases: of failing in safety trips during the
whole-operation period of the-reactors. Investigations. have shown that in some cases, for exam-
ple,in the-event of a.stop of one-of the-working circulating pumps,there.is no necessity in
dropping power to zero level, but its fast automatic: reduction up to the-level of 30% - a
so-called-second-stage-safety trip- - is quite-sufficient. In this connection apart of signals.
which previously caused power drops to a zero level has. been transferred to the-signal class
of the-second-stage ? safety trip. This has i:icreased ship vitality and ? lowered - the . number of
sharp thermal oscillations of the-plant equipment. The-automatic power drop up to a zero
level has-been referred to as "a first-stage safety trip".
All the-equipment of the primary circuits of the icebreaker's atomic power plant has
operated, including 1963, about 15,000 hours at working conditions (at a pressure -of 180 kg/cm2
and -temperatures.of 250 - 3100C). 1Main circulating pumps have'run up to 8000 - 9000 hours
without any inspection. However some - of the, pumps failed,due - to lowering the ? insulation
resistance of stator winding. Emergency circulating pump.s.have, shown reliable-operation. The
steam pressurizing system has.provided.high stability.in maintaining the-primary circuit.
pressure-at stationary power levels and. prevented pressure oscillations of more than ?5 atm
during transients.
The-active-corrosion products settling in the-lower part of the, pressurizers. complicate,
dismantling work and-electric heaters' replacement. This kind of work has not yet been
carried out. Ion-exchange-filters have-been able-to maintain the required. water quality in
the primary and-secondary circuits of the-power plant: specific resistance,of 1-2 mo/cm;
chlorine ion content of no. more than 0.02 mg p- ,r litres and pH value-of 6 - 8. Recently the
ion-exchange, resins KY-2 and AB-17 have been successfully used in filters.
The-addition of hydrodine,into the. secondary'circuit water has been often done, At the
beginning of the icebreaker mechanical plant operation several cases of a.short-term rise-of
salt concentration in the?secondary.circuit feed-water took place, But later owing to
modification of certain parts.of equipment and-control system improvement, these phenomena
have, occurred. rather seldom and accompanied by the-immediate-localization. It is necessary
to.emphasize -the -efficiency of double-grid plates in the condensers and refrigerators of over-
board water, practically excluding seawater pumping.
The'NPP steam generators.have been .reliable and-stable -under both stationary and
transient conditions. But there have-been some-cases.of their pipe-systems' unsealing in the
course, of operation. Measures for detecting and -disconnecting a damaged steam generator
were taken rather orderly and-fast. Owing to this.only'a.short-term rise in water activity of
the-secondary circuit with -the, background -levels being exceeded by several times has been
observed,
Some 'cases of drop leakage-through the-seals of main slide-valves due-to seal packing
drying have taken place-in the-primary circuits' valve system. The packing has.been replaced
by'a.higher-quality one-but it has.been necessary to tighten seals in every 1500 - 2000-hour
intervals. Tiie,sylphon valves of the-primary circuit drain system have-proved to be, insuf-
ficiently.reliable, The-annual inspection of these -valves.has.been required.
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The-actuating mechanism of the-reactor control and -safety system has shown reliable -and
stable operation and -is beyond -essential reprooves.
The , biological shield surrounding reactors, equipment and primary circuit piping lines has
proved to be-sufficient, it has. revealed no failures due-to shock loads when the-ship sailing
in ice fields. and during storms. The-highly active, slurry has-leaked-into some -pulse -tubes
coming out from the lower piping points. and;caused -rise .in gamma-radiation levels at the
places. where -these tubes.pass -beyond the,biological shield-and-go to.the-data .units of mo-
nitoring devices. and -meters. This has necessitated-to arrange-additional local shielding.
The.NPP maintenance -work has. revealed. the-necessity to expand the. sanitary treatment
block at the-entrance-to the-strict regime -zone .and,provisional stores for liquid and solid
wastes. There fore .theisanitary block and -stores, have -been -properly re-equipped. The radiation
monitors available-provided a proper control of the radiation situation on board the-ice-
breaker and, the. activity level in the atomic plant process circuits. In the-course-of plant
operation a .;.artof these -dosimeters. has. been replaced-by new improved ones. The-ice-
breaker's active-gas control system has been modified.
The, individual-control results.for the personnel operating the,atomic power plant have
shown that an integral irradiation dose -for the overwhelming majority of controlled persons
does not exceed one-third or a half the-maximum permissible dose of 5 rem cads.
Only some-persons who performed radiation-hazardous work connected with premises'
decontamination where 'the, primary circuit water leaks occurred have received doses close to
the, maximum permissible -ones (3, 4).
REACTOR CORE NEUTRON-PHYSICAL PARAMETERS
The, reactor design description is given in a report /1/. The -core, consists of 219 fuel
assemblies passing through the knots of a regular lattice with a 64-mm pitch (See figure 6);
it is.of 1,6-m height and-l-m equivalent diameter.
The, first fuel charge-of each reactor contained-80 kg of uranium-235. The-reactor core-is
designed.for 200..day_run at a:maximunu power of 90 W. One-third-of the initial fuel charge
deteriorates: during this.period. Multiplication factor original reserve-complying with this
rather high burn-up (p=.14%) has been decreased by.two-fold due?to.boron-10 addition into the
channel casing tubes :in-quantity :of 92-grams-per a:reactor for the,first fuel charge, Boron is
spread, overa:reactoc core nonuniformly: with. decreased .c:-ncentration from a.reactor axis: to
periphery; the-outer row assemblies. contain no.boron.
A fuel assembly:(see-figure,6) contains.a.36-cylindrical fuel-element .bundle-fitted.with a
spacing frame. Fuel elements :are.6.1 arm in outer diameter and minimum element spacing in
assembled bundle is 1.5.mm. Element cans. are.0.75.mm thick. Fuel assembly casing tubes,
fuel elements. cans: and, spacing frames. for the first charge were made of a zirconium alloy.
Sintered-uranium dioxide, pellets: of a.5%. average, enrichment are, used .as a fuel.
In loop test results .could -not produce -experimental information on the 'process dealt with
fuel burn-up at the-real positions of the-compensation mechanism and neutron field distortions.
Thest',data could be derived, from reactor core fuel irradiation as a whole.
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The reactor core-development was based upon calculated data. All the-calculations were
carried out with a quick-acting electron computer using two standard one-dimensional programs
(radial and high) in this case-an effective, multiplication factor and a radial neutron leakage
were assumed to be constant.
A calculation basis for the multiplication parameters in process channel lattice was an
approximation technique, founded on introducing empiric corrections derived from analysis of a
great deal of critical experiments into a classical thermal-reactor model.
Reactor operation on the icebreaker was preceeded by reactor-core, investigation at a
testing jig. During these-tests both the,reactor core design and the design of reactor control
and reactivity compensation systems were, ultimately mastered.
Figure,7 shows the-level of the-compensation system insertion into a reactor core as a
function of energy generation in reactor No.1 during its long-term operation at a power range
from 40 MW to 60 W.
The design curve.(1) is.determined from one-dimensional axial program. It is in good
consistency with experimental points corresponding to the compensation system positions
during reactor long-term operation at stationary power levels. Thus in spite, of the design
model proximity a theoretical description gives rather satisfactory results.
Fast insertion of the-compensation system at the-beginning of the-reactor lifetime is
attributed mainly to boron burn-up effect: reactivity excess caused by boron burn-up is not
compensated by reactivity loss.due to uranium-235 burn-up. The equilibrium occurs only in
79 days after reactor maximum-output operation.
The right branch of the,curve,is less crooked due-to liberation of the,fuel assembly
slightly-burned regions during the,compens;ition system removal out of the reactor core, On the
figure,7 one can see some points ~,orresponding to the reactor compensation system positions
under various poisoning at temperatures. from 400C to 80?C and maximum reactivity values
(curve 2).
During the reactor lifetime its minimum subcriticality with the, compensation system being
positioned on the-lower end switches (3) and with the control and safety rods being fully
withdrawn is equal to 1 - 1.5%. This provides. reliable, reactor blocking.
Radial energy distribution before, the,reactor operation under working condition was
determined, from the,activity'level of the-process channels exposed to.radiation within the-core
at a: temperature, about 200C. and a minimum power level. The radial peaking factor for this
case,proved-to be-1.2 (see figure-8). The design curve (see figure~9 2a2) is referred to hot-
poisoned-reactor peaking in process channels with ignoring the effect of thick-walled steel
casings.for control and-safety rods which are;located -aroundthe reactor axis. This accounts
for the divergence of calculated and-experimental data for the central process channels.
During reactor operation all the changes in the,energy-generation field shape mere, control-
led by resistance-thermometer, measurements of temperature-drops in fuel assemblies. By'the
time of reaching maximum reactivity (at an energy generation about 160,000 MW,7hr) radial
energy-release peaking had. become maximum about 1.42 (figure, 8 "b") due-to boron burn-up
in reactor central zone,
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 160,000 MW,/hr the control rod group weight rose-up to 0.7% and-was practically kept at this
level to the end of the-core lifetime.
This control rod weight increase-arises basically from extending neutron fields in the
region of their disposal and also from growing the thermal-neutron diffusion length in the.
reactor core-as nuclear fuel and-boron are,burned up.
rods, a .guide - device for placing an overloading container against a proper reactor cell,
hoisting capacity, shield in containers for unloading fuel assemblies and control and safety
equipped with a spent-fuel storage and refuelling facility including a turning crane of 12-ton
auxiliary ship "Lepse" was used to carry out this operation. The, motor ship "Lepse" is
The first overcharge,of theicebreaker reactors was performed in spring of 1963. The
REACTOR IIEFUELLING AND 13ASE MAINTENANCE
etc.
time-of one-assembly discharge was about 15-20 minutes. It took about three days to bring
a jack. The number of such wedges during reactor discharge was rather smallu.An average
In caserof an asserbly being mechanically wedged in reactor cell it is made ,a move by
icebreaker crane on board the ship "Lepse" for its loading into the spent-fuel storage,
slide of which is then closed and the container with the assembly inside is carried-by the
periscope,(6). The assembly is hoisted by the hand winch into the container, the bottom
winch to clamp the assembly head. The expansion tong operation is controlled with a
be,overloaded. After that an expansion tong (5) inside the container is lowered by a hand
[see figure 12 (1)] is put on the-shielding plate,of the, guide, device, (2) above an assembly to
The .sequence of reactor refuelling operations is. as .follows. An overloading container
The - icebreaker crane -was. also used-for reactor overcharge,
bending, , abrasion signs or notable rod diameter changes. Can surfaces are covered with a thin
they are in good condition (see figure 13). The fuel elements has revealed no swelling,
Investigation of the-fuel elements unloaded from the icebreaker reactors has shown that
about the total reactor discharge,
Figure-14 illustrates a fuel element cross ground end after a power generation level
layer of dark coloured depositions of some -nicrons thick.
temperature of uranium dioxide-that is confirmed by the absence of a central hole-in a fuel
however kept. The - temperatures in fuel core, centre - have - not been close to the, melting
In spite of the fuel element core-cracking the-space between the core and its can has been
about 15,000, MWd/tU.
being inserted, into the-reactor core-and neutron density change control being duplicated. It
During reactor charging operation fuel assemblies were pulled down with all absorbers
After removing all the fuel assemblies reactors and their primary circuits were washed off.
pillet and,an anomalous-grain growth.
consuent checking of the control rod weights and reactor shield. These new assemblies
After charging.new, modified fuel assemblies the reactors were-run critical with the
took 6-10 hours to load-fuel assemblies into a reactor core.
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were of the same geometrical sizes but manufactured Izy the improved technique. One of the
reactors was loaded with steel-encased fuel elements.
The icebreaker reactor refuelling operations were carried out in the region of Murmansk
where a moorage is arranged and the minimum equipment required for the-icebreaker "Lenin's"
maintenance is available,
TRAINING OF OPERA'T'ING FORCE
The engineering staff operating the icebreaker atomic plant went through special training at the
Makarov Higher Engineer Nautical Institute in Leningrad, After that it worked on probation at
atomic power stations. The training and practising courses on reactor control ended in going
in for examinations before the State Commission. Later on the examinations to get a working
place have been taken annually.
The common operating personnel of the icebreaker atomic plant gained the necessary
operational experience at the enterprises of atomic industry.
The first two years of the atomic power plant operation were the years of practising with the
plant. This period-enabled to reveal all the adv^ntages and disadvantages of the plant, to
outline ways for improving the, plant control system and to master the-transients. At that time
the operating personnel of the atomic power plant showed a great piece,of creative initiative,
put forward a variety of interesting proposals on redesigning the control circuits and some
components of the plant. A part of these, proposals has been already realized.
CONCLUSION
The long-term operation of atomic heat-generating plant on board the icebreaker has permitted
to make a comprehensive valuation of its working capacity under various swimming conditions.
The principal scheme and arrangement of the atomic plant have proved to be successful and
the provided reserve equipment to be,quite sufficient.
No overirradiation of the-personnel has occurred during the whole operating period of the
icebreaker. The-atomic plant has proved to be so reliable that in order to inspect the equipment
only one visit per day was required,
The first experience of installing an atomic power plant on board the icebreaker has proved to
be quite successful. The technical expediency for the building of powerful atomic icebreakers
to service the,Northern Sea Route has been confirmed,
Experimental neutron-physical. characteristics of the, icebreaker's reactor cores have been
obtained by N.A.Lasukov an&A.K.Sledziuk and their collaborators. Data derived from control
room registers and 1,)g-books have been also used in preparing this report,
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FIG.2. GENERA I. vww Or STEAM Gl?;NI;RATING POWER PLANT
1 - heat exchanger of III-IV circuit loop; 2 - pump of the internal cooling circuit; 3 - pipeline of
outside cooling circuit; 4 - air pump valves; - main circulating pump of the primary circuit;
6 - emergency circulating pump of the primary circuit; 7 - room for the thermal-control data
units; 8 - valves of primary circuit drain system; 9 -- heat insulation; 10 - primary-circuit filter
cooler; 11 - steam generator; 12 - steam generator room; 1 i - shield for the steam pipeline outlet;
14 - steam piping; 15 - emergency coolant tank; 16 - control and safety system room;
17 - actuating mechanisms of the control aid safety system; 18 - ionization chamber;
19 - carborite; 20 - reactor; 21 - reactor core; 22 - level indicators: 23 - biological shield for
the ventilating shaft: 24 - primary circuit piping; 25 - iron-water shield; 26 - equipment
storage pit; 27 - primary circuit filter; 28 - iron-water shield tank; 29 - concrete;
30 - primary circuit main pipeline shutter; 31 - piping collector; 32 - hatchway to the steam
generator room; 33 - heat exchanger of the internal coolant circuit; 34 - mechanical purification
filter; 35 - pressurizer; 36 - electric heaters.
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FIG.3. PLANT PRINCIPAL HEAT FLOW DIAGRAM
coolant circuit; SG - steam generator; P - pressurizer;
MCP - main circulating pump; ECP - emergency circulating pump;
IEF - ion exchage filter; FC - filter cooler; ICCC - internal
coolant circuit cooler; ECC - external circuit cooler; FP - feed
pump.
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FIG.4. INLET AND OUTLET TEMPERATURES AS
A FUNCTION OF REACTOR OUTPUT
(Vit --11-,
FIGS. REACTOR CORE CROSS SECTION
-1;
m
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Pa6'oquv KaHa'/7
Ton iwSN6iu 94erMeHln.
FIG.6. PROCESS CHANNEL AND FUEL ELEMENT LONGITUDINAL
SECTION
40
FIG.7. COMPENSATION SYSTEM INSERTION INTO
REACTOR CORE VERSUS REACTOR 1 POWER
GENERATION
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:i:
=
:
~
? v - -
r
r [
W U
i
n
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O
O~
~
B
?
FIG.8. RADIAL ENERGX-GENERATION
DISTRIBUTION:
?
0
e
I
FIG.9. AXIAL PEAKING FACTORS FOR THERMAL NEUTRON FIELD
AND ENERGY GENERATION AT VARIOUS POSITIONS OF
COMPENSATION SYSTEM
Experimental points:
o - at a minimum controlled power at reactor lifetime start
o - 18 MW power and 60,000 MW/hr energy generation
0 - 55 MW power and 70,000 MW/hr energy generation
? - 54 MW power and 110,000 MW/hr energy generation
3 iQ - 55 MW power and 160,000 MW/hr energy generation
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Cn,pn Trw.wnw aw A-, 'C,
FIG.10. TEMPERATURE REACTrVITY EFFECT
Experimental points during reactor warming up: o, e - outside source heating;
b - selfheating;
o - points for reactor removal
of after-heat
ono
1 A
M
M
M
MrwKT., %Y
FIG.11. REACTIVITY CHANGES DUE TO THE DOP-
PLER EFFECT
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a+
M
?
e
?
?
1 tl
40
u
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FIG.12. REACTOR WITH THE REFUELLING
EQUIPMENT
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