JPRS ID: 10150 TRANSLATION CHANNEL NUCLEAR ENERGY REACTOR BY N.A. DOLLEZHAL' AND I. YA. YEMEL'YANOV

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JPRS L/ 10150
1 December 1981
Translati on
CHANNEL NUCLEAR ENERGY REACTOR
By
N.A. Doilezhal' and I. Ya. Yemel'yanov
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JPRS L/10150
1 December 1981
ChAP~NEL P~UCLEAR E~~ERGY REACTOR
Moscow KANAL'NYY YADERNYY ENERGETICHESKIY REAKTOR in Russian 1980
(signecl to press 27 Mar 80) pp 11-78, 95-102, 119-123, 131-137, 182-203
[Chapters 2, 3, 5, 10, 11 and excer~ts from chapter 6 from book
"Channel Nuclear Energy Reactor" by N.A. Dollezhal' and I. Ya.
Yemel'yanov, Izdatel'stvo Atomizdat, 20500 copies, 208 pages]
CONTENTS
~ Chapter 2. Physical Characteristica of the Core 1
i
2.1. Structure of the Core 1
2.2. The Method of Neutron Physical C~lculation 3
2.3. Physical Experiments 8
2.4. Neutron-Physical Calculating Characteristics 11
2.5. Nuclear Safety 37 ~
.
2.6. Physical and Power Startup of Reactor 3
- Qzapter 3. Design of a Reactor Plant 45
3.1. Th e Reactor 45
3.2. The Fuzl Channel 51
' 3.3. The Reac.tor Pipelines. 58
3.4. Flow Regulators 60
3.5. Selecting the Structural Materials and the Water Q~?emical Regime.... 60
= 3.6. Thermal and Hydraulic Characteristics.......~...... 66
= 3.7. Investigating the Strength of the Equipment and Pipelines,.......... 74
i
' Chapter 5. Fuel Assemblies 81
5.1. Operating Conditions and Main (haracteristics 81
5.2. Results of Main Experimental Work to Analyze the Reliability of
- the TVS 84
Chapter 6. Channel-TyPe Nuclear Power Reactor
6.1.8. Regulation of Energy Release Distribution 88
6.2. Reactor Power Control System 89
6.4. D~mamic Processes 95
- a - [I - USSR - K FOUO]
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- Chapter 10. 1he Recharging Machine 1C4
- 104
- 10.1. Configuration and Design
Modea...... 11(~
10.2. Operating
Chapter 11. Prospects for Develop?nent of `Jr~nium-Gr~phtte Chgnnel-~pe
Reactors ~13
11.1. Princi~les far Improving the Core 113
11.2. Sectioc;al-Block Deaign of a Reactor " 121
11.3. Steam Superheating in the Core
- 11.4. 'Ihe Coolant Loop and Equipment Configuration 12~+
- b -
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PHYSICAL CHARACTERISTICS OF THE CORE
Mosc:ow KANAL'NYYYADERNYY ENERGETICI~SKIY REAKTOR in Russian 1980 (signed to press
27 Mar 80) pp 11-47
(Chapter ~ from the book "Channel-Z'ype Nuclear Power Reactor", by Nikolay Antono-
_ vich Dollezhal' and Ivan Yakovlevich Yemel'yanov~ Scientific Research and Design
tnstitute of Power Engineering, Atomizdat, 2,550 copies, 208 pages]
2.1. Structure of tche Core
The RBMK nuclear pawer reactor is a heterogeneous thermal neutron channel-type
reactor in which graphite is used as the moderator. The coolant-boiling light
water--circulates throuqh vertical channels that penetrate the lining of the core.
This type of reactox is the latest development of uranium-graphite reactors in the
USSR.
The core has the shape of a vertical cylinder 11.8 meters in diameter and 7 meters
high. It is surrounded by a lateral reflector 1 meter thick and end reflectors 0.5
uaeter thick. The core contains the fuel elements, moderator, cooiant, fuel channels
and neutmn absorber rods (control rods) (Figure 2.1).
i 4 5 6 7 B. 9 2
~ .
3�
. 2 1 .
Figure 2.1. Structure of Core: 1--graphite stackingj 2--end reflector; 3--
side reflector; 4--reflectnr cooling channel~ 5--shortened ab-
sorbing rod (USP)j 6--automntic regulator (AR) rods 7--fuel chan-
nel; S--fuel assemblyt 9--~~1 control i~) and safety system
(AZ) rod
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The graphite stacking of the reactor is 2,488 vertical columns which were assem-
bled from blo~k~ 250 X 250 mm in cross section with graphite density of 1.65 g/cm3.
The blocks have oblique openings 114 mm in diameter along the vertical axis that
are designed for location of the fuel channels and the monitoring and control chan-
nels. Graphite rods are installed in the openings of four rows of the peripheral
columns (the side reilector).
The fuel channels are located in 1,693 cells of t,he square lattice nf the core.
Part of the channel located in the core is made of a zirconium alloy and is a pipe
88 mm in diameter with wall thickness of 4 mm. Graphite rings are placed on the
pipe to ensure thermal contact with the stacking blocks. The fuel assembly, which
is two series-connected fuel assemblies (T`1S) 3.5 meters long each, is installed
inside the channel. The design gap between the TVS comprises approximately 20 mm.
The TV5 consists of 18 fuel rods which are attached by steel spacing lattices on a
central pipe manufactured of zirconium alloy measuring 15 X 1.25 mm. Either a car-
rier rod 12 mm in diameter or a carrier pipe of zirconium alloy measuring 12 X 2.5
mm passes inside it. The .fuel element is a pipe with outer diameter of 13.5 mm with
wall thickness of 0.9 mm of zirconium alloy filled with pellets 11.5 mm in diameter
of uranium rlioxide with density up to 10.5 g/cm3 and ~aith enrichment of 1.8 or 2
percent U-235. The inner cavity of the fuel element is filled during manufacture
with a mixture of argon and helium and is sealed by cathode-ray welding. Addition-
al ab$orbers (DP) are installe3 in part of the fuel channels in the initial state.
- The coolant is delivered to the bottom to each fuel channel. The economizer sec-
tion of the channel in which water is heated to ~aturation temperature is approx-
imately 2.5 meters high from the bottom of the core. The active boiling process
occurs in its remaining part wherein the mass steam content of the coolant in-
creases along the flow and comprises an average of 14.5 percent throughout the re-
actor at the output from the core.
The monitoring and control system channels are located identically to the fuel
channels in the central openings of the graphite stacking columns. The square lat-
tice for location of the 179 rods has spacing of 700 mm and is rotated by 45� with
respect to the lattice of the fuel channels. The SUZ rods are functionally divided
into groups that ensure radial control of the energy release field (the RR rods),
automatic regulation of the mean power level (AR), s~ramming of a chain reaction
(AZ) and regulation of the energy release field in height (USP). The rods of ~he
first three groups are removed upward from the core and the shortened absorbing
roas of the fourth group ar~ removed downward.
The channel for the SUZ rods is made of the same zirconium alloy as the fuel chan-
nel but has diameter of 88 mm with wall thickness of 3 mm. Graphite rings are also
placed on the outside of the channel. The absorbir_g rnds are assembled from sec-
tions of the same type, articulately connected to each other. 7."ne absorbing sec-
- tion has a sleeve design. Boron carbide sleeves measuring 65 X 7.5 ~n and with
total thickness of 984 mm ar~ enclosed in a sealed annular cavity formed by the
outer pipe measuring 70 X 2 mm and by the inner pipe measuring 50 X 2 mm, the lat-
ter of which is manufactured from aliuninur.~ alloy. The RR, AR and AZ rods are as-
sembled from five absorbing sections and have a total length of 5,120 mm; the USP
are assembled from three sections with totaY length ~~f 3,050 mm.
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A self-contained water circuit with pump-heat exchanqe unit is used to cool the
channels and the rods. Water moves in the channels from top to bottom and flows
around the outer and inner surfaces of the absorbing rod jackets, being heated from
40 to 60�C. The inner cavity of the channels in an operatinq reactor is filled
with water regardless of the location of the rods. With the exception of the AR
rods, all the rods have displacers consisting of five section~ arti~ulately con-
nected to each other with total lenqtr~ of 5,000 mm. The displacer section is a
pipe measuring 74 X 2.5 mm of altaainum alloy with sealed end caps. One section is
hollow and the remaining sections are filled with sleeves and cylindrical qraphite
blocks. When the absorbing rod is removed from the effective zone of the core, t~he
displacer is introduced into this zone and the neutron balance is improved due to
~ displacement of part of the water which is a stronq absorbp~t. The AR rod differs
from the RR and AZ rods by the fact that it has no displacer and a plate 20 mm in
diameter is installed in the lower part of the absorbing assembly to distribute the
flow rate of the coolinq water. One-third of the total coc~lant flow rate through
the channel passes through the inner cavities of the AR rods, the same as in the RR
and AZ rods.
The neutron fuel distribution through the core is monitored by a physical monitor-
ing system (SFKRE). Seven-section S-emission monituring sensors are located behind
the energy release fields in height in 12 channels uniformly distributed in the
central part of the core for this purpose. The ~-emission sensors which are in-
stalled in sealed cavities of the central carrying pipes of the fuel assemblies of
the 117 fuel channel~ are used to monitor the energy release distribution to the
reactor radius. Fission chambers, which are used to monitor the neutron �low when
the reactor is started up, are arranged in four channels lacated in the peripheral
' row of the core.
~ There are 20 vertical openings 45 mm in diameter, in which channels with three-
; zone thermocouples are installed to monitor the graphite temperature, in the re-
actor stacking in the assemblies where the graphite blocks are joined. A total of
156 channels is provided in the central openings of the peripheral row of the
graphite columns to cool the reflector. Water of the self-contained control system
~ circuit is used as a cooling medium in these channels and also in 12 channels wi.th
~ height energy release monitoring sensors and four channels with fission chambers.
2.2. The Method of Neutron Physical Calculation
' The RBMK reactor was developed on the experience of design and many years of oper-
I ation of channel-type uranium-graphite reactors of the world's first AES [Nuclear
' power plant], the Sibirskaya AES and the B~loyarskaya AES imeni I. V. Kurchatov.
' Therefore, the methods of neutron-physical calculation checked in existing reactors
were the basis for developing the methad of neutron-physical calculation of the
RBMK reactor. At the same time, due to orienta~ion toward continuoua recharging
- of the working channel.s and some design features of the RBMK reactor, suppl.ements
to the existing methods and in some cases development of new methodical solutions
were required. One can distinguish three main phases in development of the methods
of calculating the RBMK reactor: .
a) worki.ng out the physical concept of the rsactor and analyzing its main
neutron-physical characteristics on the basis of traditional methods of calculat-
ing uraniinn-graphite reactorss
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b) determining the main physical design characteristics of the reactor and
working out specialized programs for calculating the cells and lattices of the
channels of RBMK type reactors;
c) refining the physical characteristics of the reactor and making calculations
for starting it up.
Program for calculating fuel burnup in the cell. The main program which was used to
carry out a large part of the physical design calculations of the reactor and which
was used to make preparatory calculations for other programs is the VRM program for
calculating fuel burnup in *.he cpll of a channel-type reactor operating on thermal
neutrons with rod-type TVS with the geometrical dimensions and physical compc~sition
of the cell assigned. The extent of uranium burnup and the neutron balance in the
continuous or partial recharging mode, the isotope composition of the fuel and the
output of the channel as a function of time with given geometric dimensions, the
initial enrichment of the uranium, the average output of the channel and neutron
leakage beyond the cell are determined in the program. The fuel and moderator tem-
perature is determined in the corresponding program blocks. Besides the main ver-
sion, the program has several modificatons: calculation of a sin~gle-row lattice,
claculation of reactivity effects and calculation of input data for heterogeneous
programs and two-group constanta.
The r.equirements o~' speed and simplicity of the algorithm determined the selection
of simplifien calculating methods. A channel with coolant temperature and density
averaged in heiqht is considered. Neutron breeding in the high-energy zone and ab-
sorption on U-238 resonances are calculated by ordinary methods for a homogeneous
lattice. A Wigner-Zeitz cell is distinguished when considering the processes in
the thermal energy zor~e, but neutron overflow is introduced on its boundary, which
has different signs for fresh and burnt-up channels and provines an integral bal-
ance of neutron generation and absorption during the run. It is assumed that only
thermal neutrons leak beyond the cell. Neutron capture in the epithermal energy
zone is assumed to be low. The neutron spectrum in the thermal and epithermal
zones is represented in the form of the sum of the Maxwell spectrum with effective
neutron temperature dependent on cooxdinates and the Fermi spectrum and the West-
cott cross-sectior~ system is used in this case. Neutron flow is calculated in P1-
approximation for the model of a two-zone cell. The thermal neutron flux distri- .
bution through inidividual fuel elements of the assembly is also calculated in �
P1-approximation. The presence of non-breeding channels in the reactor is taken
- into account by introduction of effective spacing (graphite "smearing").
- The GE heterogeneous program that takes into account epithermal effects. The most
- common method of calculating the characteristics of a heterogeneous lattice of
- working channels is the heterogeneous method [2, 2]. In the classical heterogene-
ous method th~ reactor is regarded as a syatem of working channels placed into a
moderatinq medium, while channels of finite dimensions are regarded as filamentary
sources of fast neutrons and sinks for thermal and resonance neutrons. The ele-
mentary diffusion equation is used to describe diffusion of thermal neutrons be-
tween channels. The absorbing properties of the channels are characterized by the
logarithm derivative of the thermal neutron flux to the channel surface.
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Application of the classical heterogensous m~thod to calculations of modern power
reactors is very difficult. First, the channel volume comprises a significant
part of the cell wlume so that ~he assumption of their filamentary nature leads
to an appreciable error. Second, a considerable number of neutrons is slawed down
inside the channel, as a result of which it is a source not or.ly of fast, but of
thermal neutrons as well. Moreover, a considerable part of the neutrons is ab-
sorbed and generated in the epithermal energy zone in power reactors ~ith high
uxanium burnup with accumulation of isotopes having cross-sections with high reso-
nance integrals. Modification of the heterogeneous method, that permits one to
take into account the i.ndicated characteristics of the channel lattice, has the
arbitrary name GE. 7.'he breeding coefficient and the neutron flu~c distribution
through the cell channelg with periodicity having a maximum of 16 heterogeneous
channels can be calculated by the GE program. The GE program is used as a block in
the program for calculating the fuel burnup of the channel lattice with regard to
the height heterogeneity of properties.
The heterogeneous burnup calculating program. The method of calculating fuel burn-
up xn a channel-type reactor operating in the continuous recharging mode.(with het-
erogeneous core) is based on two assumptions. First, the structure of the core
charge and the sequence of recharging the working channels are assumed such that
the core is a periodic lattice of fuel channels with different fuel burnup, the
main repeating element of which is the feed-rate cell (or macrocell). A macrocell
consisting of 14 fuel channels and two control channels is usually considered when
calculating an RBMK reactor. This assumption simplifies calculatior: of the neutron
, fields through the macrocell, but it is suspected in this case that the considered
macrocells are located in an energy release distribution plateau (or the reactor is
j infinitely large i,n radius). Second, it is assumed that the :nacroscopic constants
i of the fuel channels are determined anly by the extent of fuel burnup in a given
~ channel and are not dependent on the properties of adjacent channels. Zn this case
j calculation of burnup in a heterogeneous system is broken down into two independent
parts:
i
~ a) calculation of the macroscopic properties of fuel channels as a functian of
; the degree of fuel burnupt
I b) calculation of power distribution through the macrocell, cansisting of dif-
� ferent channels whose properties are determined in the first part of the
~ calculation.
I
i
The GE program is used in the latter part as the main block. The power distribu-
tion obtained in this manner is assumed to be fixed during a given time step, which
can be selected as sufficiently small. This separation of burnup calculation per-
mits one to significantly reduce the calculating time since the channel properties
are not calculated directly in the second part but only a sample is taken from the
properties.
The capability of distinguishing up to eight zones in height (the HINDI program) is
provided in the program to take into account the inhomogeneity of the �uel channel
properties in height that results from variation of the coolant density, nonuniform
fuel burnup or variable fuel charge through the length of the channel. The neutron
flux distribution is calcu].ated in single-group appraximation. The program takes
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into account variation of the core charge structure accarding to one of the channel
recharging schemes and the determined reactor operating n~odes, for example, when
the multiplication factor decreases below a given value ox when other conditions
are fulfilled: upon reaching a given deqree of burnup and with channel output. It
is ~ufficiently universal and permits one to investigate different fuel charges of
~hannel-type reactors. Unlike the ~i program, the real structure af the reactor
feed-rate cell and the presence of nonmultiplying channels in the lattice are taken
into account in it. The limitations of the program capabilities are related to
representation of the core in the form of a set of identical macrocells and with
the assumption of constant coolant density distribution during the run through the
height of the channel, although the nature of the energy release distribution var-
ies considerably, especially during the initial period of reactor operation.
Programs for two-dimensional calculation of a reactor. The development of computer
technology cansiderably expanded the capabilities of realizing calculating methods
which could not be accomplished previously. Attention was primarily devoted to
those problems of reactor calculati~n which had been hardly investigated, specif-
ically, on calculatiora of the energy release distribution through the entire reac-
- tor. The experience of operating the reactors of tl~e Beloyarskaya AES indicates
that, unlike urani~.un-graphite reactors operating on natural uranium, local surges
of energy release determined by the local inhomogeneities of the core structure can
occur in reactors with slightly enriched fuel. A program for two-dimensional cal-
culation (in x-y configuration) of the energy release fields was developed for the
M-220 computer for reactors of the Beloyarskaya AES with approximately 1,000 chan-
nels [3~. The numerical method of calculati.ng the criticality and the energy re-
lease distribution used in it was realized in two-dicnAnsional programs for up to
3,000 channels with the appearance of the BESri-6 computer. The BOKR program was
used extensively to process the results of physical startup of the reactor and to
form a full charge. The BOKR-COBZ program [4~ is now used as the operating program
for calculating the energy release fields after introduction of additional units
into the program that take into ~ccount nonuniform contamination of the fuel with
xenon and fuel burnup.
Solution of a system of two-group diffusion equations for a reactor consisting of
= various types of square cells is realized in the BOLR program. The physical ~rop-
erties are constant inside each cell. The assemblies of the calculating grid coin-
cide with the centers of the channels. Upon comparison of the calculated and ex-
perimental data, it was shown that this arranqement of assemblies was more
preferable than in the angles of elementary cells of the reactor as was used in
[3], The iteration process, in which the neutron source distribution and the new
neutron flux distribution of the first and second groups are taken into account as
a result of sequential bypass of all the assemblies of the calculating grid on the
basis of given initial single neutron flwc distribution in each energy group and
for the initial value of the multiplication factor equal to one, was used to solve
the system of diffusion equations. The next iteration was then carried out using
the neutron fluxes found as a result of the previous iteration. The new value of
the multiplication factor is calculated after each iteration.
This iteration scheme is one with "mixed" iterations in which, i:nlike the classi-
cal scheme with external and internal iterations, the neutron sources are correc-
- ted after each iteration rather than after some approximation of the neutron flux
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for them. This scheme was convergent and convenient by the rate of calculation.
The method of sequential Young and Frankel overrelaxation used to accelerate the
converqence of internal iterations in the classical scheme, was used to reduce
- the calculating time.
- The presence of c~ntrol rods and other nonbreeding channels in the core is taken
into account by assigning the corresponding homogenized properties of cells in
which these absorbers are located. Partially inserted rods are replaced by fully
inserted rods equivalent to the first ones in efficiency.
A method of heterogeneous calculation of a reactor that extends the classical het-
erogeneous method to reactor systems having a large number of cells (up to 2,500)
was ~eveloped in [5, 6l. A heterogeneous reactor is represented in the form of a
finite lattic:e of. thin rods (channels) in a infinite moderator. Neutron transport
in the moder.~ator is described by two-group diffuaion equations of the Galinin-
Feinberg type using the quasi-albedo method. mao-dimensional diffusion equations
with filamentary so:lrces--sinks--are approximately rewritten in a form similar to
- the finite-difference method. To do this, the reactor is covered with a square
grid equal to the spacing of the lattice so that the sink-sources are located on~y
at its points.
The QUAM program (its subsequent modifications are NEWQUAM and QUAM-2), which per-
mits calculation of the energy release distribution to the channels of a react~r
whose core is inscribed into a sguare with side of not more than 48 cell~s, was com-
piled by the indicated method. The number of different varieties of rods (channels)
does not exceed 99. There is the capability of assigninq individual characteris-
tics for each channel; linear intierpolation is made for all channel parameters ~when
determining the dependence of properties on burnup.
It was shown upon comparison of the results of c~lculation by the BOKR and QUAM
programs that both programs yields sufficiently close energy release distribution
through the reactor channels.
The programs used in physical calculations. Without dwelling on the characteristics
of the programs, let us present- a list of the names and designation of programs used
in makinq physical calculations of the RBNIIC reactor and intended for making tradi-
tional reactor calculation~.
The SI-5, P3-15 and ?=3-50 are programs for solving the single-speed kinetic equa-
tion in P3-approxima~tion. They are used to calculate the thermal neutron flux dis-
tribution thzrough the elementary cell. ~
The DOP and MOV are programs for solving single-dimensional two-group diffusion
equations. They are used when calculating the energy release f~elda in heiqht and
radius of the homogenized reactor model.
POIS is a program for calculating transient poisoning. It is used to calculate
reactor poisoninq in transient modea.
VOR is a program fnr calculatinq fuel burnup "at the point." It is used when oal-
culating the physical properties of cells with chan~nels differing from the model
used in the VRM program.
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DSP is a two-group synthetic program for two-dimensional calculation of a reactor
by the separation of variables method. ~ao-dimensional calculation is replaced by
sequential one-dimensional calculations.
GRIM is a program to determi.ne the physical characteristics of nonbreedinq channels.
"FUGA" is a modernized heterogeneous program for calculation of c'riannels with re-
gard to fast, resonance and epithermal neutron effects determined by the presence of
nonbreeding channels in the charge.
~ "FIALkA" is a program for calculation of corrections to heterogeneous constants of
working channels detexmined by the presence of nonbreedinq channels near them.
2.3. Physical Experiments
The compl~xity of the core structure of a reactor operating in the c~ntinuous re-
charging mode, the presence of channels with significantly different breeding, ab-
sorbing and moderating properties and the larqe dimensions of the core made f~sll-
scale experiments essentially impossible due to their high cost and mainly due to
the long periods required to conduct them--development of a full-scale bench, manu-
facture of channels, assemblies and so on. However, the need for an experimental
check and refinement of the methods of calculatinq complex iattices was clear.
Therefore, experimpnts w~~re conducted during the period of working out the design
on an insert of an already existing qraphite bencr;~ the spacing of the channel lat-
tice in the insert was equal to the desiqn spacing and comprised 25 cm and the
number of channels was 81s the height of the investigated systems was 3.5 meters,
i.e., it was one-half the design height. The experimental data ware used to correct
the calculating methods and to analyze the characteristics of the reactor.
Experiments to determine the characteristics of complex lattices were conducted in
' phases--from simple to complex. Homogeneous lattices of RMBK-type TVS consisting
of rod fuel elements of uranium dioxide with natural U-235 content were investigated
during the first phase. The lattice multiplication parametQrs and also the neutron
fields in the cell were measured. It was shown ir_ these experiments tr.at the sys-
tem cannot be critical for an RBNIIC-type lattice when using uraniuzn di~~xide of natur-
al enrichment. Therefore, further experiments with natural uranium were conducted
in a subcritical assembly surrounded by a seed region. The experiments were con-
ducted at 25�C with a dry zone and one filled with water. The positive nature of
reactivity upon dehydration of the TVS channel lattice was show:l.
The neutron distribution through ::he channel height in the rupture zone was mea-
sured and the degree of increase of neutron flux density was determined ~due to the
structural rupture of fuel through the core height between TVS. It was shown that
this surge drops rapidly as the distance from the channel axis in~reases so that
introduction of a displacer in the center of the TVS leads to appreciable equa].iza-
tion of the neutron field through the channel radius. The results of ineasur.~~n~ents
made with assemblies of 16 fuel elements with two percent fuel enrichment are pre-
sented in Table 2.1. The qap between the fuel in the fuel elements comprised 41 mm
along the vartical and the material of the end parts was SAV alloy. Analysis of
experimental data made it possible to conclude that the surge of energy xelease on
the peripheral fuel elemer.~.:. of fresh TVS of the RBNIIC reactor comprises 35-40
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percent in the operating state and decreases exponentially as the distance from
the gap increases with relaxation length of 1.3 cm 1.
Table 2.1. Relative Surge of Neutron Flux Density in Region of TVS Rupture
Material Between Fuel Elements
SAV and H20 SAV
PoinL of Measurement Inner Ring Outer Ring Inner Ring Outer Ring
In center of rupture 1.83 1.58 1.51 1.39
At fuel boundary 1.53 1.37 1.35 1.25
Far from rupture l.a 1.26 1.0 1.13
Besides experiments with homogeneous lattices, experiments were conducted with
mixed lattices which were assembled from assemblies of two percent and natural
enrichment. The mixed lattices were formed of polycells which were squares con-
sisting of four channels. The types of investigated lattices and the results of
determining their materials parameters ~ 2 with and without water in the fuel chan-
nels are presented in Table 2.2.
, Comparison of the values of ~t~ for the considered lattices with and without water
in the fuel channels showed that the effect of variation of reactivity upon dehy-
dration of the channels is negative for a lattice containing a two percent enrich-
ment assembly and positive for the other investigated types of lattices. With
~ average enrichment of 1.7-1.8 percent U-235 thsough the polycell, the effect of
' dehydration is close to zero. due to the dependence of ~2 on the average enrich-
ment, it was concluded that the miniuntan enrichment at which a cold nonpoisoned
reactor with water in the fiuel channels can be made critical comprises approxi-
mately 1.2 peresnt.
i
Table 2.2. Values of Mate rial Parameter a�
Gfiarge Averaqe Enrichment 2~l~-~4 ~ 2 ~ 2= ae~ without
I of Through Polycell, water- 2 with
_I Polycell percent Without Water With Water water, 10-4 cm 2
~ p;4 0.714 -0.712 -3.87 3.16
~ 4:0 2 5.74 6.43 -0.69
3:1 1.67 4.57 4.07 0.5
2;2 1.35 4.07 1.67 1�~~.
1:3 'Ll 1.42 -1.03 2.45
2:1+
+1 empty** 1.57 3.12 2.12 1
*The ratio of the number of enriched assemblies to the number of assemblies with
natural fuel compositiQn is shown.
**The effective spacing was 33 cm ~
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As one would expect, the neutron spectrum in lattices with enriched fuel is more
rigid compared to that in a spectrum of assemblies containing natural uranium.
- Filling the fuel channels with water leads to si~nificant spectrum softening and
varia~ion of the neutron field through theSi~Se~creasesebya15y20 percenteleThetex-
of the outer ring in which the neutron den y
cess of effective neutron temperature at the cell boundary over graphite temperature
comprises 160 and 100�C, respectiv~ely, for assemblies with enriched fuel with and
without water and comprises 110 and 80�C for assembliQS with natural oxide fuel.
The coefficient of nonuniformity of neutron density through the assembly, deter-
mined as the ratio of the maximum density in any fuel element to its average value
through all fuel elements, increased by 3 and 5 percent when the channel was filled
with water for channels with natural and enriched fuel, respectively, but did not
exceed 1.1 in a sin~le case. The dependence of DP effectiveness on the ratio of
sleeves of two percent boride steel ("heavy" sleeve) and ordinary steel ("light"
sleeves) was studied experimentally in different systems, but mainly with and with-
out water in fuel channels having Z'VS and in DP channels. The fraction of epitherm-
al absorption n= PeP/(p~~ + pDp) for DP in a channel without water was equal to
3.0 percent for a l~ght DP, depending on the ratio of heavy and light sleeves; it
was 1-5 percent for DP with ratio of heavy to light sleeves of 3 and was 8 percent
for heavy DP. Here pDP and PDp are the effectiveness of absorption in the epi-
thermal and thermal regions of the spectr.um, respectively. Taking the sleeve de-
sigh of the DP into account, the effect of internal water on the absorptivity of
the DP was investigated, which showed that the internal water increases the eifi-
ciency of the DP. For example, the efficiency of a heavy DP increases by 13 per-
cent with the presence of internal water. Thus, the experiments showed that effi-
ciency decreases, all things being equal, when the DP is dehydrated an~? it
- confirmed the feasibility of introducing an interrial aluminiun displacer.
Experiments on the effect of an outer water layer on the compensating capability
of a DP with internal aluminum displacer showed that a layer of water around the
DP decreases its efficiency by 10 percent. This result is true of both a heavy DP
and of a DP having boride and ordinary sleeve ratio of 1:1. However, if the chan-
nel with absorbent consisting only of light sleeves is filled with water, its
efficiency is increased.
The same experiments were conducted with absorbing rods similar to reactor control
zvds having boron carbide as the absorbent. The experiments showed that the pres-
ence of water inside the rod increases its efficiency by approximately 5 percent,
while the outer layer~ of water considerably reduces its compensating capability.
Thus, the efficiency of the rod is reduced by approximately S percent with layer
thickness of 2 mm, by 9 percent with thickness of 4 mm and by 13 percent with
thickness of 6 mm. The fraction of epithermal absorption with water inside the
rod compz~ises 18-20 percent.
- It should be noted that all the experimental data were found in small critical
assemblies which only simulated th~ different fragments of a full reactor charge
and different conditions for a cold nonpoisoned reactor. Therefore, they were
used to check and correct the calculating methods.
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- 2.4. Neutron-Priysical Calculatinq Characteristics
2.4.1. Main Periods of Reactor Operation
The neutron-physical characteristics of the RBNBC reactor were calculated for three
main operating periods: The first phase of reactor operation with initial charge,
the so-called transient period which precedes the steady recharging mode and the
steady continuous fuel recharging mode. Each of the indicated periods has its own
specific direction. The reac~or shoulc~ have the best engineering and economic in-
dicators that ensure competitiveness compared to electric power plants of other
types in the steady fsel recharqinq mode. The initial reactor charge should pro-
vide dependable compensation of the initial excess reactivity of the fresh fuel
with optimum engineering and economic indicators, The transient period is charac-
terized by continuous variation of the core structure and its neutron-physical
~ characteristics. The main calculated neutron-physical characteristics of the RBNIIC
reactor and their variation as a function of the reactor operating period are con-
sidered below.
2.4.2. The Initial Reactor Charge
Versions of accomplishing the initial charge and the passage of the transient per-
iod were consisiered: reducing the fuel enrichment in the assemblies of the initial
charge, using assemblies with different fu~l enrichment, incomplete core charge and
~ so on. One or anc~ther version was selected on the basis of the neutron-physical,
~ heat engineering, e~:onomic and dynamic investiqations of the core and of the in-
~ stallation as a whole. Problems of optimizing the initial fuel charge are consid-
~ ered in more detail in [7-9], the maia results of which are presented in Table 2.3.
Different compositions of th~ periodicity cell consisting of 16 cells, including
two SUZ channels, are presented in it.
It follows from Table 2.3 that the local misalignment of channel output in the
~ periodicity cell for all versions except version 2 appreciably exceeds the corre-
sponding value equal to 1.25 for a steady reactor operating mode. This means that
for the thermal loads not to exceed the values used for the steady mode the reactor
should operate at reduced power for som~: time. The same thing can be noted with
regard to nonuniformity of the field in height. The coefficient of axial energy
release nonuniformity not exceedinq the value in the stead~ mode, in which the en-
ergy release is equalized in height as a result of fuel burnup, can be provided
~ only in versions 2 and 6 by selecting the corresponding properties of the DP in
i
height.
i
' The principle of an incomplete reactor charge and introduction of the DP into the
freed cells was used on the basis of investigations when forming the initial fuel
charge of the RBMK reactor. Selecting the length of the DP and of the distribu~ian
of its absorbing properties in heiqht is determined by two factors. First, the DP
rods should provide compensation of reactivity (together with rods of the control
system) in all states of the r8actor. Second,. th~ DP rods should contribute to
equalization of the energy release field through the reactor height to the required
limits. The DP composition was corrected directly during physical startup of the
RBMK reactor: the heavy and light abaorbinq rings were assembled in a ratio of
1/2 in the upper and lower sec~ions 1 meter long .each and in the ratio 3/1 in the
central section 5 meters long.
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a
0
.r., v,
~ ~
rl C Rf N ~ ~ ~ ~ 01
i~~l ~I fn ~ CI I~ r l~ f\ l`
9 b ~ 3 ~ w ~ ~ ~-~I ~ ~
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w r-+ ~ ~
o~o
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~ N N ~L'j r-I
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V
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m
N
~ �i-1 N 0~ O N r-I N ~ N f"1
b~ ~ ~ ~
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~ o 0 0 0 0 0 0 0 0
�~-1 ~ O O t0 N ~ O ~1 O ~7'
CT ~ c~1 ~1' O~ ~ O 'C n M
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~k1 ~ r-1 N N ~ N ~ r-1 ~ r-I
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'-I M~~1 V' ~"1 V~ P~ V~ C r-1
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H z ~3 > H
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~
a*
a~ -
~
U i~ �
�r1 �rl i~ ~ 'l7
W.~ ~ Op tr1 l~ (1D DO 01 y
d i) U . . . . . . ~
U ro k r-I ii rl ~-1 ~-1 O N
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Selecting the location of installing the DP in the core. Since the DP rod is in-
terchangeable by design dimensions with the assembly, locating them in the core
is not limited by any design solutions and different methods of locating ~he DP
with respect to the SUZ channels in the periodicity cell can be considered. Ca1-
culations of the periodicity cells by the GE proqram with different arrangement of
~ the DP and different content of SUZ channels showed that the multiplication factor
may differ in this case by approximately one percent w3thout rods and by ~.5 percent
with SUZ rods, while the maximum output of the fuel channels may differ by approxi-
mately 2 percent. Of greatest interest is comparison of two symmetrical arrange-
ments of the DP in the periodicity cell with respect to the SUZ channels, shown in
Figure 2.2. The locations of the DP were compared with full initial charge of the
reactor in the working state both with and without SUZ rods. With tihe standard
arrangement of the DP, the control rods regulate the output of the freshly~charged
assembly more effectively than with checkerboard arrangement. Z'his circumstance
forces one to give preference to standard arrangement of the DP since both ver-
sions differ slightly in efficiency of the DP and the SUZ rods.
. qn
~ qn ~
qn
~ ~ ,qn
a ~
' Figure 2.2. Arrangement of DP in Periodicity Cells: a--standard; b---checker-
~ boardj the control channel is noted by the symbol
' Energy release distribution through core height and radius. The distribution of
~ the absorbing properties through the length of the DP h,~ve a significant effect on
i the shape of the axial energy release distribution during the initial moment of
"i reactor operationt the SUZ rods may have the same effect. The actual arrangement
i of the control rods in the core in the absence of programs for three-dimensional
i calculation of the reactor can be taken into a~count only approximately. This type
i of analysis shows that the shape of the field may vary very slightly in height
when the rods are shifted and the USP rods must be located below the center of the
~ core to compensate for the field distortions of incompletely inserted RR rods.
I The nonuniformity coefficient in heiqht can be maintained in the ranqe of 1.25-1.50
with selected composition of the DP.
! Calculated investigations of the energy release distribution through the reactor
radius can be arbitrarily divided into three phases. The ratio of the numbe~ of
DP in the central and peripheral homogenized regions of the reactox required for
given equalization of energy release through the coxe radius is determined in the
~irst phase. The specific cartogram of ~P arrangement in the core is refined dur-
ing the second phase by means of two-dimensional programs for calculating the re-
actor as a whole. The physical startup period of the reactor when a larqe number
of calculations is made by two-dimensional proqrams, from the results of which the
initial charge of the reactor is selected with regard to the data actually
achieved in the reactor, should be taken into account by th~ third phase of
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investigating the radial energy release fields. (The initial charge contains
1,450-1,440 assemblies and 230-240 DP for operating RBMK reactors as a function of
the production tolerances on the charge an3 enrichment of the fuel and c~raphite
density). It was shown by calculations ~hat a pos~tion of SUZ rods can be selected
in the hot poisoned state of the reactor in which the r~activity compensated for by
them is approximately equal to the design operational reserve of reactivity, while
nonuniform energy release distri.bution thzough the radius comprises 1.28.
Calculated investigations of a full charge and late~ the experi:ence of operating
the RBMK reactor showed that its characteristic feature is the high sensitivity of
neutron fields to displacement of the control members. This is related to the fact
that the high excess reactivity is compensated for by a large number of absorbers
" and when some of them are removed (especially the peripheral absorbers) a region
occurs, sometimes close to criticality and containing 15-20 channels with TVS among
which there is not a single absorber. In this regard the point of locating the
rods and the RR rods to be removed should be selected very carefully, observing a
specific sequence of removing th e RR rods.
Based on numerous calculations to equalize the energy release fields in the reac-
tor, it has been suggested that all 8~ RR rods be divided into four groups as a
function of their location in the reactor (Table 2.4). The fourth group combines
the peripheral rods and the central rods are divided into three regular lattices
embedded into each other. Operational compensation of the excess reactivity is ac-
complished at each moment of ti.me by the RR rods of one of the central groups and
by the peripheral rods which are shifted eo as to equalize the currents of the side
ionization chambers. The rods of each central group are shifted sequentially, oc-
cupying an approximately identical position in height with deviation of + 0.5 rneter
from the mean position. The rods of the two other central groups occupy the ex-
treme upper or lower positions depending on the reactivity reserve. The indicated
procedure for removing the RR rods permits a radial nonuniformity coefficient of
approximately 1.8 to be maintained.
Table 2.4. Division of RR Rods by Groups*
I rpynne 1I rpynna ill rpynna I IV rpYnno 1 rpynne I,li rpynne il[ rpynna IV rpynne
I ~
22-21 16-25 22-25 12-31 42-41, 36-45 42-35 36-65.
22-31 16-35 22-35 12-35 42-51 36-55 42-45 42-11
22-41 16-45 22-45 12-41 42-61 4fr-15 Q2-55 42-65
_ 22-5~ 26-t~ 22-55 12-45 52-21 46-25 6a---25 4f~-I1
32-21 26-25 32-15 16-21 52-31 46-35 5~35 4&-65
32-31 26-35 32-25 16-55 52-41 46=45 52-45 52-I5
32-41 26-45 32-35 22-15 52-51 46-55 52-55 56-21
32-51 26-55 32-~~5 22-61 521-61 56-25 62-35 56-61
32-61 36r--15 32^-55 32-11 62-31 56-35 62-45 62 35
42-21 36-25 42-15 32--65 62-41 ~_31
42-31 36-35 42-25 36-I1 62-5! 56-55 66--35.
- , 66-41
� 66-45
*
Arrangement of the rods in the reactor is ahown in Figure 2.3.
IGey :
_ 1. Group
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When reactor power is reached at which the readings of the sensors of the physical
monitoring system become reliable, the field is ac~ually equalized from their read-
ings by using rods of different groups.
Effects and coefficients of reactivity for initial r~actor charge. Determinatien
of the temperature and density ef�ects of core reactivity of the RBMK re3ctor is
one of the most complex calculations. This is largely determined by the complex-
ity of the core composition and also by the essentially total absence of experi.men-
tal data by the time of reactor startup which w~uld make it gossible t~ correct the
methods of calculating the effects of reactivity. Experimental data were obtained
on an insert or 81 channels and on critical assemblies having half the core height
where rods 2 or 3 meters long were used rather than on a full-scale bench. In this
regard the calculations of reactivity effects were corrected and were partially
done again with regard to the experimental results obtained in the reactor.
1D l1 14 /6 ?0 ?2 2f 26 30 S? 3f ~6 40 ~f? 4i S6 50 52 5f 56 60 62 6f. 66 '
67 . �
ss
~ n n
� �~n �~n �~n sy
- 63 n n
� 62
61 n n n n
.n...n ..n.. 60
57 n n n n n n
_ ~ 56
55 n n n n n
...n... . 54
53 n n ^ " ^ "
52
5~ .n n n n n n n n ~ 50
�~fl � � �~f1
y~ n n n n n n n n n
45 n n n n n n n�
�.n �s� y4
, 43 " ~ 42
tir ~ ~ ~
- yp.
~ ~ ~ n n n
36
,~j n n
...n ~
n n ~
32
31 ~ ~ n
�~n ;~p
Y7 � n n n n n ~ 26
15 n n n n n n
� 4
43 ' n n n n n n n
- 22
2/ n n n n n n n n n
. ...n 2(J
17 n n n
f6
15 n n
..n ..ri . n ..n (4
y !3 " 12
~
. !0 .
1f tQ ~-1 ~-d p-5 ~-7 x-9 ?-11 Ol ~ E'S 67
a-2 6-0 n-8 o-f0~
x
Figure 2.3. Initial Reactor Charge of Second Unit of Leningrad AES: 1--RR
rod; 2--USP rodj 3--AR rodi 4--AZ rodt 5--AZ recompensating rods
. 6--energy ~release sensor in height (DKE (h)); 7--starting fission
chambert 8--additional absorber1 9--uncharged channel; 10--fuel
assembly with energy release sensor through radius (DKE (r))j
11--fuQl assembly
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Three main p�roblems werz considered for the initial charge: variation of reactiv-
ity upon dehydration of the core in the cold etate, variation of reactivity upon
heating of. tha~ care and determination of the reactivity coefficients in the wr~rkinq
state. Th~ calculated and experimental data on variation of reactivity with dehy-
dration of the core of a cold unpoisoned reactor are presented in Table 2.5.
Table 2.5. Effec~ of Dehydration
Dehydrated Channels Calculation Experiment
~ With TVS -0.00055 -0.001 + 0.0005
With DP -0.00726 -0.011 + 0.0013
SttZ 0.0141 -0.010 + 0.0013
Table 2.6. Calculated and Experimental Temperature Coefficients of Reactivity,
_ 10_SoC_1
Estimate from Experimental
Reactivity Coeffic;ent Calculation Data
aE -4.8 -5
a~ 3.4 3
a~v + at -8.5 -8
Study of the variation of reactivity with heating o= the core made it possi.ble to
analyze such important reactor characteristics as the temperature coefficients of
moderator reactivity a~, water at~, fuel at and total temperature coefficient aE
The calculation was made for the temperatuxe ranqe of 100-200�C on the assumption
that 90 control rods wsre inserted into the core (Table 2.6). The reactivity co-
efficients were calculated fo.r r~ominal working valuea of heat engineerinq param-
eters of the reactor on tr;~ assumption that 30 fully charged rods w~re located in
its core in this state. The calculated values of the reactivity coefficients were
as follows: a~ = 0, at~ _-~.1�10"'SOC-1, at =-1.0�10'S"C-1; the density coeffi-
cient of reactivity was aY = 1.44�1G'2 cm3/g and the steam coefficient of reactiv-
ity was a~ _ ~P/0~ _ -1�10'z.
Variation of the reactivity coefficients during reactor operation is considered
be low.
Neutron flux density distribution through the assembly and fuel elements. The
thermal flux density distribution through the cell was calculated by the SI-5 pro-
gram in P3-approximati,on with regard to thermalization for a homogenized model of
the channel. For conveience in making heat engineering calculations, the flux
density distribution through each fuel element, according to [10], is reduced to
the form
~A; (r,0) _ ~p, q~r cos A) exp (~ir2),
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where pi is the relative flux density level in the fuel element of the i-th row,
qi is the flux density gradient through fuel elements of ~he i-th row, ~i is the
coefficient of cArrosion of the flux density in the fuel element and 9 and r are
the polar coordinates of a point with respect to the center of the fuel element.
The neutron d~.stribution through the microcell, consisting of a fuel element sur-
rounded by an equivalent amount of water, was calculated to determine parameter
_ in P3-approximation. Parameter which was equal to 0.2 cm'2, was determined by
the neutron distribution inside the fuel element. It was assumed that the value
of ~ is identical for all fuel elements. Parameters pl = 0.611, ql = 0.0331 cm 1,
g2 = 0.706 and q2 = 0.0845 cni 1 were determined for fuel elements of the inner
(i = 1) and outer (k = 2) ring by the neutron fl+ux density distribution throuqh the
cell. As a result the following expressions were found for energy release in rel-
ative units: .
~D, (r, 0) _(0,611 -I- 0,0331 r cos 6) exp (0,2 r'~; ~1=0,6~5;
m2 (r, U) (0,706-}-0,0845 r cos 6) exp (0,2 r~); ~Z = 0,765; 0,73,
where ~1, ~2 and ~ are the mean values of the volumetric energy release in the
fuel elements of the inner ring, the outer ring and in all the fuel elements,
respectively.
According to the derived expressions, the coefficient of nonuniformity of energy
release through the fuel elements was Kt~ = 1.05, the coefficient of nonuniformity
of the specific energy intensity of the fuel was KV = 1.11 and the maximum coeffi-
cient of nonunifc~rmity through the radius of the fuel element was Kr~ = 1.06. The
calculation corresponds te uniform distribution of fissionable isotopes through
' the radius of the fuel element. Accumulation of Pu-239, which will occur to a
~ greater degree in the ou~er layers of the fuel element core, distorts the form of
; energy release in the fuel, but the considered distribution is the most dangerous ~
for maxi.mum fuel temperature.
i 2.4.3. The Transient Operating Period of the Reactor
i
i The transient operating period of the reactor, i.e., the time from the initial
charge to the stationary fuel recharging mode, is characterized by continuous vari-
ation of the reactor parameters and the composition of the core. One of the most
I significant problems during this period is that of organizing continuous recharging
! of the fuel assemblies tha~ satisfy the conditions of maintaining ~he reactivity
II reserve and nonuniformity of energy release within given limits. Solution of this
' problem can be complicated by additional operating conditions, for example, delay
~ of the deadlines for introducing the loading-unloading machine, the need to re-
charge specific channels and so on. Recommend~tion on the sequence of recharging
the DP were presented for conditions of planned continuous recharging and the con-
sumption of assemblies, the extent of burnup of removed fuel and other indicators
of the reactor required to determine the economic characteristics of the plant
during the transition period were determined.
For convenience in calculations, fuel recharging can be related to periodicity
cells. In this case the core was divided into two radial zones: a central and
- peripheral and it was assumed that the periodicity cells in each zone are
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ir7entical and have their own constant total output datermined by the number of
cells in the zone and by the average output of the zone. The calculated periodic-
ity cell in the central zone has a mean output of 28.9 MW and contains 12 fuel
assemblies, 2 DP and 2 SUZ channels at the initial moment. The calculated cell in
the peripheral zone has a mean output of 21.8 MW and cor.tains 13 assemblies, 2 Dp
and 1 SUZ channel.
All the periodicity cells in each zone are recharged by an identical program with
specific sequence of assembly replacement. This sequence of channel recharging is
maintained throughout the entire operating life of the reactor. Channels arranged
_ identically in the periodicity cells are recharged to maintain the charge symmetxy
and its periodicity in each reactor zone. It is obvious that under rea~l condi-
tions the recharging program can be corrected on the basis of the actual power gen-
eration of the channels and the energy release distribution through the reactor.
Moreover, disruptions of the adopted sequence may occur on the periphery of the re-
actor where identical periodicity cells canno~ be clearly determined.
' Calculating the fuel recharging mode. The results of calculating the recharging
mode by the HINDI program are presented in Tables 2.7-2.9. According to the cal-
culation by this program, the average extent of fuel burnup in the steady recharg-
ing mode comprises 19 GW�day%t, which is in agreement with the planned value of
18.5 GW�day/t within several percent.
The data presented in Tables 2.7 an3 2.8 permit one to estimate the operating con-
ditions of the initial charge channels in the reactor and to find several values
that characterize the transient operating period of the reactor, for example, the
consumption of assemblies for the entire reactor during the transient operating
period (Table 2.10). The consumption of fuel assemblies comprises 475 with the
reactor generating 106 MW�day of thezmal energy in the steady recharging mode.
Effects and coefficients of reactivity. Due to the complexity of the core struc-
ture and the nonuniform distribution of burnup through the core, calculation of the
effects and eoefficients of reactivity during transient operation of th~ reactor is
_ ~ a complex problem. Therefore, they were designed with average burnup of 5 and 10
GW�day/t of the charged channels initially since removal of the DP through one of
each periodicity cell is completed with burnup of 5 GW�day/t and removal of all DP
is completed with burnup of 10 GW�day/t. The calculated vaiues of' the reactivity
coefficients and the indicated moments of the run are presented in Table 2.11.
The results indicated in Table 2.11 were found on the asswnpt~San th4t the reactor
has nominal heat engineering characteristics and that 20 control rods are inserted
into the core to compensate for the operational reserve of reactivity.
Misalignment of output during recharging of the DP. Calculations of a reactor
in the DP recharging mode presented above and of replacement of the DP by fuel as-
sembZies were made on the basis of the condition of simultaneous replacement of
DP in all periodi.city cells of the reactor. In this case only the power redistri-
bution inside the periodicity cell is taken into account. In fact, periodicity
cells with different number of DP are located in the reactor, which distorts the
neutron flux distribution through the reactor.
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Table 2.7. Main Characteristics of. Recharging Mode in Central Zone
(1~ Bpewa Ao neperpyaKr, Caybexa adropaxxa, ~3~ ileperpyao4-
}{aMep neperpy3xx ~2~ s~~. cyr I'Br�cyr~'r utiA xo~~~x�
B f19lAK! OG- ~ ~6~ /7` ANlNT
` ` l /
PMOAM4BOCTM 4/ 5) ~
N~-
M~6
MNMN~0A6HOC I 11~]IC~NaJIbXO! MNNMMIA6N~f1 ~ CQlJ~XAfI MiKCMMBAbHaA KIIEp
~
1(nep erpysKa . 100 3~0 0 0~ 0 1,182
p,II)
2(ne erpy3xa 390 630 0 0 0 1,085 ,
~m ~
3 630 670 11,8 ~12,2 ,12,6 1,193
4 670 720 12;6 13,0 13;4 1,193
5 720 ?70 13,4 13,8 14,2 1,193
6 ~ 770 620 14,3 14,7 15,3 1,192 '
7 820 890. 15,0 15,5 16,0 1,132
8 890 95b 16,1 16,6 17,1 1,193
9 950 1020 17,1 17,6 18,2 1;191
10 1020 1100 18,1 18,7 19,3 1,185
11 1100 1180 19,2 19,5 20,5 1,175
12 1180 1260', 20,3 20,6 20,9 1,153
13 1260 . 13401 20,6 21,1 21,6 1,142
~ 14 l340 1430 ~ 21,4. 22,2 23,0 l,llf3
15 1430 1510 ~ 22,3 22,5 22,6 1,149
16 1510 1590 18,8 20,0 21,2 1,162
- 17 1590 1660 17,2 � 18,1 19,0 ~ 1,182
18 1660 1740 18,2 18,5 18,8 1,18t '
- 19 1740 ' 1810 18,9 ' 19,0 19,1 1,184
' 20. 1810� 1890 19,1 19,3 ~ 19,5 1,180
21 1890 1970 19,4 19,6 19,7 " 1,184
22 1970 2050 ' 19,7 19,9 20,0 1,185 '
23 2050 2130 , 20,0 20,2 20,4 1,184
24 2130 2210 2U,1 20,1 20,1 1,206
~ 25 2210 2280 , 19,8 19,9 20,0 1,174
26 2280 236b ' 19,7 19,8 19,9 1,182 .
i 27 2360 2430 19,1 19,2 19,3 1,176
28 2430 2510 19,0 19,1 19,2 1,156'
; 29 2510 - 19,5 1,141
I Key:
- 1. Number of recharging anc~ periodicity cell
2. Time to recharging, eff. days
~ 3. Extent of burnup, GW�day/t
' 4. Minimum ~
I 5. Maximum
. 6. Average
I' 7. Recharging factor Kper = N~g/N ~
~ 8. Recharging of DP
The experience of calculating the reactor recharging by two-dimensional programs
shows that first, the misaliqnment occurring upon replacement of DP by a~ssemblies
are greater than would follow from calculations of the periodicity cells (the co-
efficiezt of misalignment reaches 1.5) and aecond, moving the control rods permits
a significant reduction of misaslignments, bringinq them up to the accepted design
values.
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Table 2.8. Main Charac~eristics of Racharging Mode in Peripheral Zone
Ta 6xxs sdropauxa; Tfepecpyaoa-
~1~ Bpc~e Ao neperpy:+Kx, t3~ y r.wll ~o3~~rx�
Howep neperpya- 2 cyr � TBs�cyr~r ~q~ext -
tcx e AaeAKe - 5. . (6) ~7~ N~tca
nepxoAxaeocrx ~ 4 naD .
MIfNMMH~bH~E I ~(2%CMMlJ16NOt ~6HHXYlAbNlN ~ CQlAHNA' Y9RCfl~(~J16H8A K N
1(rte 8~pysKa 50 430 0 0' 0 1,119. ~
~m ~
2(ne erpysKa 430 T90 0 0 0 1,073
,~n~ ~
3 ' 790 840 10,6 11,0 11�,4 1,i98
4 840 900 11,4. 11,7 12.0 1,197
5 g00 960 12,0 12,4 12,8 1,i99 �
g 960 1030 12,6 13,2 � 13,6 1,197
7 l030 1100 13,3 13,4 13,5 . 1,192 �
g 1.100 1170 14,0 14,3 14,6 1,180 ,
g 1170 1250 14,3 14,8 15,3 I;164
~p 1250 1330 15,3 15,7 16,1 , 1,143 '
1~ 1330 1410 15,1 16,5 16,9 ~ 1,122 �
12 1410 1500 16,8 17,3 1'7,8 , 1,114_
13 . 1500 1600 17,7 18,2 18.7 l, l l l
14 16b0 1.700 18,8 19,2 19,6 1,110
15 1700 lgp0 19,6 20,1 20,6 1,098
~g 18U0 1910 20,6 21,1 21,7 1,129
17 1910 2000 16,8 , 18,0 19,2 1,111�
18 2000 20'90 15,5 16,4 17.3 1,155
- lg 2090 2180 . 16,8 16,9 17,1 1,20�
20 2180 2280 17,1 17,2 17,5 1,208
21 2280 2370 17,5 17,6 . 17,7 1,209
22 2370 2460 17,1 l'F,2 17,3 1,210
23 2460 2550 17,3 17,4 17,5 1,190
Key :
, 1. Number of recharging and periodicity cell
2. Time to recharging, eff. days
3. Extent of burnup, GW�day/t
4. Minimum
' S. Maximum
6. Average
7. Recharging faator Kper = Nm~s/A
= 8. Recharqing of DP
2.4.4. The Steady Fuel Recharging Mode
The extent of fuel burnup. The chaxacter.i~atics of the RBMK reactor in the steady
continuous fuel recharqing mode are presented in Table 2.12. A reactor with
height-averaged properties is considered: average coolant density of 0.516 g/cm3
and average coolant temperature of 28~�C. The operational reserve of reactivity
is equal to 1 percent and leakage to the height is 1 percent, which corresponds to
a reactor model with unequalized power distribution by height. Correction for
self-equalization of power by height during burnup is introduced into the average
burnup of removed fuel to the reactor. With regard to the data preaented in
Table 2.12, the average fuel burnup through the reactor is equal t~ 16.5 GW�day/t
and the averaqe run of the channel is approximately 1,100 effective days. The
fuel removed from the reacCor contains on the average, kg/t:
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Table 2.9. Isotope Composition of Fuel, kq/t
~b~ ' � ' (2)
Rwropaxxe, ~U � n~Pu iOPu ~U ~U ~pPu ~Pu 1[Iaaitx
fBr�cyr~r
0 17,870 0 0 982,13 0 0 � 0 0
1,825 15,763 0,9216 0,0053 980,96 0,3432 0,0640 O,OOOl2 . 1,9452 .
3,635 13,937 1,5393 0,0335 979,79 0,6374 0,2108 O,OA185 3,8655
5,405 12,335 1,9094 0,0827 978,63 0,8913 0,3888 0,00760 5,7341
7,122 10,916 2,1703 0,143Q 977,50 1, I l22 0,5730 0,01899 7,5406
8,774 9,662 2,3357 0,2062 976,38 1,3037 0,7555 0,03668 9,2734
10,358 � 8,556 2,431'4 0,2673 975,29 1,4692 0,9317 0,06060 10,9280
11,871 7,579 2,4795 0,3237 974,23 1,6124 1,0978 0,09058 12,5Q6
I3,320 6,714 2,4956 0,3743 973,19 1,7362 ' 1,2518 0,12590 14,008
14,702 5,948 2;4906 0,4186 972,18 1,8432 1,3931 0,16600 15,438
16,023 5,270 2,4719 0,4567 971,19 1,9355 1,5217 0,21025 16,800
17,286 4,668 2,4448 0,4892 970,22 2,0150 1,6382 0,2580G 18,098 ,
18,495 4,136 2,4129 0,5166 969,27 2,0832 1,7433 0,30889 19,337
- 19,655 3,663 2,3786 0,5396 968,3} 2,1415' 1,8379 0,36225 20,52 l
20,768 3,244 2,3436 0,5586 967,43 2,1912' 1,9227 0,41773' 21,656
Key: ~
_ 1. Burnup, GW�day/t 2. Poisons
. U-235 4.1 U-236 2.1
Pu-239 ' 2.4 Pu-240 1.7
Pu-241 0.5 Pu-242 0.3
- U-238 969.3
Energy release distribution through the height and radius of the reactor. The en-
ergy release distribution through the height of the core was calculated by the
HINDI program with regard to nonuniform water density through the length of the
channel, nonuniform fuel burnup and the actual design of the control rod displacer.
It was assumed that displacers are located in both SUZ channels in the periodicity
cell. The nanunfform fuel burnup coefficient through the length of the channel is
equal to approximately 1.4 and maximum burnup of approximately 22 GW�day/t is lo-
cated at a height of approximately 5 meters from the bottom of the core. The non-
uniform energy liberatinn coefficient througn the height bi zn~ ~i�u~ ieacLVr
equal to 1.3 in the steady mode. The maximum field is located at a height of ap-
proximately 4 meters.
. The enerqy release field through the core radius can be represented in some cases
in the form of an equalized field having a plateau zone and a peripheral zone on
which local misalignment from the SUZ rods and other inhomogeneities in the core
are imposed. The equalized neutron flux deneity field through the reactor radius,
calculated by the one-dimensional two-group program DOP for a reactor model with
five radial zones, is shown in Figure 2.4. The coefficient of field nonuniformity
through the radf us is Kr = 1.1. Local d.istortions characterized by an irregular-
ity factor Klok~ Whose value, according to the experience of operating uranium-
graphite reactors, is equal to 1.15, must be applie~? to this field. Thus, the
- maximum channel output (without regard to the recharging coefficient Kper) ex-
ceeds the average channel output in the reactor by a factor of approximately 1.2~.
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Table 2.10. Consumption of Fuel Assemblies During Tran~ient Period as a Function
.of Power Generation
E, 106 MW�day ~ E, 10~' MW�day ~ E, 106 MW�day ~
~ 0 1.5 154 3.0 815
'0.25 6 1.75 187 3.25 950
O.S 30 2.0 225 3.5 1,075
0.75 57 2.25 375 3.75 1,195
1.0 87 2.50 520 4.0 1,325
" 1.25 120 2.75 670
Note. E= is the integral generation of thermal power by the reactor;
NT is the thermal output of the reactorj and g is the cons~nption of assemblies per
charge of the reactor
~ Table 2.11. Calculated Values of Reactivity Coefficients
Average Fuel Burnup in Reactor, GPJ�day/t
- Coefficient 5 10 j
I
a, 10'S �C 1 3~2 5.4
C
a , 10"5 �C-1 0.42 5.0
tv
a~, 10'5 �C-1 _1.0 -1.1
t
aY~ I~-2 ~m3/9 -0.22 -1.30
q,~~ lp-2 0.15 0.92
v teD ~
~1) � OB
_ Y~ ~ ,
. E ~ KaNana~ oxna~rdcHU,~ ompuaramena
CFu,o
Q�4
~~.y lI n
pu~epuuNaa ~6n, cmn '
4 m
, c c a? I nePu~epuuHap o6naemb
_ ~ Odnacme naamo 5
E '
Q 100 200 d00 400 5A0 600 ' ~40'
Paduyc, cM .
Figure 2.4. Neutron Flux Density Distribution Through Radius of Reactor
(Kr = 1.1)
_ Key:
1. Ne.utron flux density, relative uni~s 5. Plateau zone
2. F~eflector cooling channels 6. Graphite
3. Second peripheral zone 7. Radius, cm
4. First peripheral zone
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Table 2.12. Main Characteristics of Reactor in Steady Fuel Recharging Mode
`1, ~2~~~:xa (cM. pt+c. 2.~)
XapaKrepecrxxa ~~o ~ nepx~epxrl- 11 nepnQ~e-
~A~ 118A 5,~pxAxaA
Y
PaAxyc aoNa, cM (6) ' 465 541 612
KonxvecTao: (7) .
A4eeK tg 1087 384 413
A9CCK dep~0AN9HOCTH ( ) 68 24 26
3arpy~lceHH~x Kaxanos ~~Q) 952 336 405
HC38I'p}~MCENHNX K8H8~110B Ql],~ ' 13rJ 46 8
CPCAHHA MOII(HOCTb K8H8A2~ KB7' {12, ]~80 1940 ~47~ '
CPCIIHN~ K03~(pHI~NCHT p33MHOMC2kNS 1,02 1,05 1,07
ITJIOTHOCTb ROTOK2 TCA~110H61X N8~l7'POFiOS:~~4~
xa rpaxxue Ave~xx, ~Q~~ C~'['P.~~CM~�C~, (1 ) 8,37 7,85 5,17
e ~ronnxee, orH. eu.: I16~ ,
a Hayaae xaMnaNex (17~ ;0,591 0,585 0,580
H KOHI(C K2Mi18HHN �:0,570 0,564 0,559
B BOAe, OTH. CA�: ~~.9T
e xayane KaMnaeNN ' ~ 0,639 0,686 0,681
B KOH42 KaMiIAHHH , 0,661 0,656 0,652
e 3aMCllnNrene, oTH. eA.: (20)
e eayane KaMnaHNx ' ~ 0,973 0,969 0,967
~ B KOHL(C KAMI7.7HHN ~ ~,941 0,942 ~~942
; K03(p(pNl(HeHT ~123MH0?KCHNA:~21~ '
s xawane xaMrta~NH 1,203 , 1,204 1,207
~ s xoHUe xaMnaxNH 0,832 0,685 Q,920
` MouiHOCrb Kat+ana� xBr: (22)
~i e Nayane x~MnaH~iH ' 2540 2365 1745
B KOHLIC K8MI18HHN 1~405 1450 1140
~ B~ropaxNe TonnHea' I'Br�cyr/r (23) f 21,13 . 18,42 16,65 '
KaMnaHex xaxana. ~~14) r � 1160 1030 , 1230
~ NsoTOne~i~ cacraa a~rpyHCaeworo ~ronaxea, xr/r. (2 )
` ~ssU ' � 2,88 3,94 4,72 .
` s~U 1 ~ 2,26 2,12 2,02
~ xuU ~ 967,1 969,3 970,1
I saop~ , ' 2,20 2,29 2,36
s+op~ , 1,99 1,76 1,58 ,
s~~pu . � ' 0,50 0,46 � 0,41 �
I wnaxt+ (26) 22,04 19,28 17~47. .
-i *
The data are presented without regard to nonuniform burnup through the absorption
heiqht in SUZ channels and the heterogeneous eff~ct. .
Key:
~ 1. Characteristic 15. At cell boundary, 1013 neutrons/
! 2. Zone (see Figuze 2.4) (cm2�s)
~ 3. Plateau 16. In fuel, relative units
I 4. First peripheral 17. At beginning of run .
5. Second peripheral 18. At end of run
6. Radius of zone, cm 19. In water; relative units
7. Number 20. In moderator, relative units
8. Of cells 21. Multiplication factor
9. Of periodicity cells 22. Output of channel, kW
10. Of charged channels 23. Fuel burnup, GW�day/t
11. Of uncharged channels 24. Channel run
12. Average output of channel, kW ' 25. Isotope composition of removed
13. Average multiplication factor ~fuel, kg/t
14. Neutron flux density 26. Poisons
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The recharging coefficient ICPer characterizes an increase of output of a freshly
charged channel compared to the mean value. According to calculation Kper = 1.21
and the autput of a freshly charged channel in the plateau zone of the RBMK re-
actior may reach approximately 3,000 kW with regard to the coefficients indica~bd
above.
The output of each production channel in the reactor must be known to solve a
number of design and operational problems, specifically to determine the reliabil-
ity of the core and to select the distri.bution of coolant flow rates through the
core channels. This calculation was possible with dev~elopment of the BOKR and
QUAM programs for two-dimensional calculation of the reactor. If they are used,
it is no longer necessary to divide the core into zones, while the maximum output
of the channel is determined by the actual state of the core and by the position
of the control members. Calculations of the radial power distribution by two-
dimensional programs confirmed the possibility of achieving an irregularity fac-
tor of Kr = 1.27.
- Reactivity effects and coefficients. The reactivity effects and coefficients in
the standard fuel recharging mode were calculated on the assiunption that there are
channels in each periodicity cell of the reactor with burnup of 0, 1, 2, 4, 5, 6,
- 8, 10, 12, 14, 16, 18 and 20 Gw�day/t. The avexage fuel burnup is equal to 9
~ Gw�day/t. Assuming that 2Q SUZ rods are complete3y inserted into the core, the
following reactivity coefficients were found: a~ = 5.2�IO'5OC 1, at~ = 4.9 10'5OC'~,
aY 2.14�10'2cm3/g, aT =-1.15�10'5oC'1 and a~ = 1.52�10-2. The effect of de-
hydrating the working channels in the cold state for a reactox with 25 rods re-
moved was also determined for the steady fuel recharging mode and comprises 1.1
percent.
Natural indicators of the fuel cycle. The natural indicators of the fuel cycle
such as consumption of enriched uranium, constunption of natural urani.~:m, accumula-
tion of secondary nuclear fuel and so on are usually considered to compare the en-
gineering and economic parame~ers af the RBNIIC reactor to those of other types of
reactors and to investigate fuel cycles. The natural indicators of the fuel cycle
of the RBMK reactor are presented below:
Initial charging of reactor
i~iass oi cnargeci uranium, zons iv~
Initial enrichment, percent 1.8*
Enclosure of natural uranium, tons 515
Transient operating mode
Leng~h of transient period, effective days 1,400
. Average extent of uranium burnup of initial
charge, GW�day/t 16.5
Average consumption of enriched uraniinn, t/year 38.0
Enrichment af uranium charged during transient
period, percent 1.8
_ Average consumption of natural uz'anium, t/year 96.0
Average productivity of reactor in fissionable plutonium
plutonium isotopes, kg/year 172
Average content in removed fuel, kg/t:
U-235 5.0
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Pu-239 and Pu-241 3.0
- Steady recharging mode ~
Mass of charged uraniuan, tons 192
Initial enrichment of loaded fuel, ~rcent 1.8
Extent of burnup of unloaded fuel, GW�day/t 18.5
ConsLUnption of enriched uranitmt, t/year 50.5
Consumption of natural uranium, t/year 136
Plutonium productivity of reactor, kg/year:
all isotopes 253
fissionable isotopes 146
Content in unloaded fuel, kg/t:
U-235 4.1
Pu-239 and Pu-241 2�9
Graphite mass in core, tons 2,000
Mass of zirconium pipes of channels, tons 103
Mass of zirconium in assemblies, tons 74
Constiunption of zirconium, t/year 19.3 ~
Note. Analyses o� the operational data and additional calculations showed the
possibility and feasibility of increasing the fuel enrichment. Conversion for
enrichment of 2 percent is now carried out. Conversion to enrichment of 2.4 per-
cent and higher is planned in the future.
~ The effective fraction of delayed neutrons A significant fraction of the output
' in the RBMK reactor is generated as a result of fission of Pu-239 nuclei aacumu-
j lated during operation, the fraction of delayed neutrons of which is considerably
lower than that of U-235: sU_235 ~ 0.0065 and Bpu_239 = 0.0021. The effective
~ fraction of delayed neutrons Beff decreases in this regard as the extent of fuel
~ burnup increases (Table 2.13). For a reactor brought up to steady fuel recharging
mode, Seff = 0.0045.
Table 2.13. Variation of Effective Fraction of Delayed Neutrons
~1~Be~ropexHe, I ~ ~ ~ Betropaxxe. I ~N BmropeHae, ~~4
r9r�cy,rrr ' rBr�cyr/r ~ rsr�cyr~T
q
� 1 ~
0~ 0,0065 7,13 0,0046 13,2 0,004U
1,00 0,0060 8,08 0,0045 14,0 0,0039
I 2,10 O,OOb6 9,00 0,0044 14,8 0,0038
3,15 0,0054 9,90 0,0043 15,5 0,003T
4,1 T 0,0051 . 10,80 0,0042 16,2 0,0037
5,18 O,OObO 11,60 0,0041 . 16,9 0,0036
6,17 0,0047 12,4 0,0040 1T,6 0,0036
Key:
l. Burnup, GW'day/t
Repoisoning of reactor upon variation of output. Variation of reactivity as a re-
sult of xenon poisoning during variation of output of a power reactor is of qreat
= interest. This is related to the fact that one or another reactivity reserve
should be maintained as a function of the nature of load variation in the power
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system and of the corresponding requirements placed on variation of reactor power,
which has a direct effect on the extent of fuel buxnup. Al1 the characteristics
o~ the RBMK reactor were calculated on the assumption that the operational reactiv-
� ity reserve is equal to 1 percent. In this case a reduction of reactor power to 50
percent of the nominal level is possible without it falling into the iodine pit.
The operationai reactivity reserve must be increased to expand the range of per-
missible variation of power, which either reduces the extend of fuel burnup or re-
quires an increase of the initial fuel enrichment to maintain the extent of burnup.
The number of control rods that compensate for the operational reactivity reserve
also varies. These functions are presented in Figure 2.5. The operational reac-
tivity reserve also affects the permissible time of complete shutdown of the reac-
tor or a reduction of power or on the ti.me of scramming of the reactar if it falls
into the iodine pit. Thus, if reactor power varies from the 100 percent level and
if the operational reserve is 1 percent, the permissible time for complete shut-
down of the reactor is approximately 1 hour and the forced shutdown time is approx-
imately 24 hourst these times are equal to 3 and 18 hours, respectively, for an
ogerational reserve of 2 percent.
~ F~ . , . . . .
4~iD0 ~~19 ~,1' ~4~ ~ ' 2 . .
~ v aE ~ ~ 3 ~ .
q f
~ ~ ~~'1B ~ 1,8 g . .
ox 5D a ~ ~23
Q3 b
~ iy z !~T � . . .
. 0. 0 Q5 . 1,Q .1,5� Z'~' 2'S'
. . ( 5 ) o~qomnud~ni savwa paa~rmuBtinanu,~
Figure 2.5. Dependence of Reactor Characteristics on Operational Reactivity
Reserve: 1--extent of burni~p with initial enrichment of 1.8
percent; 2--initial enrichment with burnup of 18.5 GW�days/t;
3--permissible outputs 4--number of RR rods inserted into core
_ Key:
- 1. Permissible power of reactor, percent 4. ivumber oi ruc rods, unizs
2. Extent of burnup, GW�day/t 5. Operational reactivity
3. Initial enrichment, percent reserve, percent
Cases of cyclic variation of station load over 24 hours were also investigated:
16 hours at 100 percent power and 8 hours at reduced power (Figure 2.6). This
operating mode was permissible for an operational reserve of 1 percent with dump-
ing of load up to 50 percent, but reactivity over a period of 24 hours varies by
approximately 1.2 percent (by approximately 0.8 percent toward a decrease and 0.4
percent toward an increase from the steady poisoning level during operation at
constant nominal power). The operator should compensate for variation of reactiv-
ity by the RR rods, providing minimum power misaliqnments in this case.
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. ~
t\ . \ 5 � .3 .
E100 654 3 4 .
~ _ ~ ~x 4 . 6
3 �
' g 30 a 3 -
2
1 .
~ 0 2 4 6 8 10 f2 f4 16 1B 20 2224 20 2 y 6 d 10 f2 N 16 fe 20 222f
BPEMA ~ 9 ~ ~2 ~ BPEMA~ 9 ,
; ~ .
Figure 2.6. Daily Schedule of Reactor Power Variation (a) up to 0(1),
20 percent (2), 40 percent (3), 50 percent (9), 60 percent
(5), 75 percent (6) and the corresponding variation of the
degree of xenon poisoninq during cyclic nperation (b)
Key:
- 1. Output, percent 2. Time, hr
Some functions of intesest during operation, related to reactor poisoning, are
presented in Ta.blas 2.14-2.16.
Table 2.14. Variation of Degree of Xenon Poisoning OK$e on Time During Operation
o� Reactor at 100 Percent Power
~ Bp ~nl,, AKX~. '~i ~ 'BDaif~. � ~~Xa~ x Bp~A, a _ OKXe. �~i ~ . .
~
0 0 12 1�,93 24 ' 2;68
2� 0,2T 14 ' 2,13 26 2,74
4 0,66 16 2,29 . 28 � 2,79
6 I,OS 18 2,42 30 2,82
~ 8 1,4Q 2tj ~ 2,53 ~ 2,98
10 ' .1,69 ' 22 2,61 . .
Key:
1_ TimP. hr
Table 2.15. Variation of Xenon Poisoning With Total Reactor Shutdown, Percent
. Motq~oen peucropa nepep ocranoa-
~ OTpas (eu~ e xo0, !~f.
.100 '76 I 60 ~0 2b
CTauHOn$pHOe ~3~ (4) 2,98 2,80 2,b3 2,36 1,94
MBKCNMBAbHOC, B NOAHOA nMe 6,35 b,05 3,73 3,20 2,24
I~JIy6NH8 NOAHO~I AMH ~C,~ 3,37 2,25 1,2 0,85 0,30 ,
Key:
_ 1. Poisoninq 2. Reactor power prior to shutdown, percent
[Key continued on following page]
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[Key continued from preceding page]:
3. Steady 5. Depth of iodine pit
4. Maximum, in iodine pit
Table 2.16 Steady Xenon Poisoning at Different Levels of Reactor Power Np
Np~ %I~KO ppp, y~I~KXe.
/o N i.
l0 1.15 60 2,66
20 ' 1.74 70 2,77
~ 30 2,11. 80 2,85~
40 2,35 90 2,92. ~
50 2.53 100 2,98 '
The nature of variation of xenon poisoninq in time with total shutdovm of the re-
actor is presented in Figure 2.7 as a function of the previoue power level.
9
- 8 ~q
~ NP-150'/. .
120 .
6 100
,
80 ~
~5 : ' e
60 , e
k~` 4 - E
a ~ , ~ -
3 ~4
~ ZQ y
2
2) ~
- .
0 10 10� 30 �?U �5U Du ~
. (1~BP[MA~ 4
Figure 2.7. Variation of Xenon Poisoning Aft~r Shutdown of Reactor Operating
at Different Power Levels
Key:
1. Ti.me, hr 2. Fuel burnup, GW�day/t
Effect of deviations of p~oduction parameters on reactivity and extent of burnup.
The dependence of burnup and of some other characteristics of a reactor on fuel
density, coolant density, channel output and other parameters for the reactor
plateau zone is presented in Table 2.17. The initial value of the mean multipli-
cation factor of the core in the steady recharging mode is equal to 1.03 (1.01 is
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th~ o~erationel reactivity reserve and 0.02 ia roactivity for leakAge throuqh the
height of the reactor with reqnrd to equalization of the neutron flux durinq burn-
up); fuel density is Yt a$.814 g/cm3, water density is Yv = 0.516 q/cm3, graphite
density is Y~ = 1.67 g/cm3, average channel output is 2,100 kW and uranium en~ich-
ment is C~ = 17.87 kg/t. The estimated dependence of the extent of burnup and the
- initial multiplication factor on fuel enrichment is presented in Figure 2.8.
1,d
a
i
~ ~ �
~ ~~2 .
b
~ ~
i .
7 ~ ' ~ ~ ,
~ .
~ 2 �
� 1,2 ,
a
b
< ~ ~
e ~
T '
tl
x
~ "7 .
~ � .
~ 1~S 'Q~g
i ( 2) Oboaau{eNUe monnusa, y. . '
-I Figure 2.8. Dependence of Initial Multiplication Factor (1) and Extent of
Burnup (2) on Fuel Enrichment (lattice spacing of 26.7 cm)
i Key.
I 1. Initial multiplication factor 2. Fuel enr3chment, percent
I ~`F~~inM~j ~f n_n~~n1 anrl Mnn~tnrin~ Ms+y_nhpra
~
~ The control and monitoring members of the RBMK reactor consist of 179.absorbing
I'i rods. They are functionally divided into 89 manual control rods (RR), 12 auto-
' matic con~rol rods (AR), 21 emergency power reduction rods (AZ) and 21 shortened
absorbing rods (USP). The SUZ channel has an outer diameter of 88 and inner diam-
eter of 82 mm and is made of a zirconium alloy of mark 125. The abaorbing material
of the rod ie BqC with density of 1.65 q/cm3. The parameters of the absorbent, mm,
are:
Outer diameter of aleeve 65
Inner diameter 50
outer diameter of jacket 70
Length of absorbing section:
of RR, AZ and AR rods 5,120
of USP rods 3,000
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Table 2.17. Effect of Production Deviatians on Reactivity and Extent of Burnup
~1~ tny6b'a N3 tx~xe jj~~~b'+dp H3~7'ex~xxe
rny6xxa h09~1�M4NEHT . H898lIbHOi'O .
IlapaMerp sdropsxxn, Q~~pBHNA K03~~H4NlH7~
TBT�Cy7/T j'Bt�CyT/T P83MHOMClHHA pa3NHQMClNNA
Oceoaxoe corroAxxe ~6) 21,017 - 1,211615 ~
Ns?~e,xexxe nnon+ocTx mnn3~e~ Ao 21,118 0,101 1,212293 0,00068
8,914 rJcM'
Ns?+exexxe nnoTxocrx eoA~?~8~ 1[0 21,640 0,623 . 1,215447 0,003832 ~
0,416 r/cM~ ~ ~ rQ~ �
I'I3M@HCHNC K03~~Hi(Ne11T3 P891lHOxiett~ta 21,920 . 0,903 1,211468, -0,000147 ;
AO 1,02 , ~
N3MeHexxe cpeuHe~ uouiHOCrs xa~ 21,022 . 0,005 1,212628 0,001013
Ao 1900 xBr
YisMeHexxe nnomoctx rpa~x~'L~Ao 20,912 ~-0,105 . ' 1,~12645 0,001030 ~
1,72 r/C~c' ~12 ) .
NsMexeHxe o6orau~exxa ypa Ao 21,921 ~ 0,904 1,220405 0,00879
18,3~ xr/r � ~
YMexbwexxe xapym~oro AxaHe~rpa~lx~~- 22,240 ' 1,223 : 1,231459 0,019844
Naabxoii ~rpy6d Ao d,.D~84 YK r1 l �
YeenNyeaxe Macc~ cranx a xa~xAd~K$- ~ 20,152 .-0,865 1,198020 -0,013595
xane xa 1 xr . ~
YKeHbmexNe TOAWHHd o6onowK~~ 21,131 0,114 1,214540 0,002925
na xa 0,1 Me+ npN coxpaxexNx e~yrpeH-
iiero ANa~erpa o6onoKxx ~i~ 21,522 ' 0,605. ~ 1,216661 0,005046
YuepbweHNe rbnutNx~ o6onovxt~"'?~~� �
na ea 0,1 ~KU npx yBCJ11iReFINH Hxyrpex-
Hero ateawrr a o6onovKe �
Key:
1. Parameter
2. Extent of burnup, GW�day/t
3. Variation of extent of burnup, GW�day/t
- 4. Initial multiplication factor
5. Variatiun of initial multiplication factor
6. Main state .
7. Variation of fuel density to 8.914 q/cm3
8. Variation of water density to 0.416 g/cm3
9. Variation of multiplication factor to 1.02
10. Variation of average channel output to 1,900 kW
11. Variation of graphite density to 1.72 g/cm~
12. Variation o� uranium enrichment to 18.37 kg/t
13. Decrease of outer diameter of channel pipe to dnar - 84 ~
14. Increase of mass of steel in each channel by 1 kg
15. Decrease of thickness of fuel jacket by 0.1 mm while maintaining inner
diameter of jacket
16. Decrease of thickn~ss of fuel jacket by 0.1 mm with increase of inner
diameter of jacket
5 mm long each
Structurally the rods are made of individual absorbing sections 967.
with gap of 65 mm between sec~.ions. The rods are cooled by a special water cir-
cuit witih temper~ft74emmf~6d length ofe4960Amm to reduce theaharmfullabsorption of
outer diameter
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- neutrons in the cooling water. Thus, if these rods are completely removed from
the core, a displacer is located in the channel symmetrically with respect to the
center of the core while sections of the channel 8pproximately 1 meter long from
the top and bottom of the displscer are filled with water. The AR rods have no
displacer and the channel is completely filled with water when they are removed.
The travel of the RR and AZ rods is 6,250 mm.and that of the AR rods is 4,500 mm.
= Being located in the extreme upper poaition, these rods are completely removed
from the core and are separated by 200 mm from its upper boundarY. The USP rods
have travel of 7,000 mm and are completely removed through the bottom of the core
and in this position the upper end of the USP rod is located at the level of the
lower boundary of the core.
The efficiency of the individual atructural components of the SUZ channels in the
reactor plateau zone are essentially independent of the reactor operating period--
initial or steady. The efficiencies preaented in Table 2.18 were calculated with
respect to a solid graphite block for an abeorber equal in length to the height of
the core.
Table 2.18. Efficiency of Idealized SU2 Rod, 10'4
State of Working Channels
- Contents of SUZ Channel Hot Cold with Water Cold Without Water
Absorbing rod 8.12 6.08 ~�69
~ 0.75 0.69
Displacer 0.45
' Water column 2.64 3.31 3.36
II The neutron flux distribution through the height for a steady recharging mode is
; taken into account to determine the efficiency of rods with regard to their real
dimensions and arrangement through the height of the core. The curv~ of the rela-
tive effectiveness of the absorbent corr~sponding to this flux is shown in Figure
2.9 as a function of the depth of insertion. The reactivity introduced with com-
~ plete displacement of the RR, AR and USP rods located in the reactor plateau zone
~ is presented in Table 2.19. The distribution of reactivity through the rods of
~ different types is presented in Table 2.20.
~ Ta.ble 2.19. Efficieny of Real SUZ Reds in Plateau Zone, 10'4
~ Stata of Working Channels
' ~d Hot Cold With
Water Cold Without Water
RR, AZ 6.27 4.34 5.50
AR 3.55 1.78 2.79
USP 4.49 3.54 4.53
The values of the total efficient of the rods in the reaotor presented above were
calculated for a two-zone reactor model through the radius (the periphery plateau).
Therefore, the efficiency of the rods was al.so calculated by two-dimensional pro-
grams. The calculations show that if the peripheral rods whose efficiency is de-
termined to a significant degree by the specific structure of the surrounding zone
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are excluded from consideration, then the ~fficiency of the rod~ is proportional
to the square of the neutron field in ~ta~ xegion of insertion of the rod with ac-
curacy up to 10 percent. Therefore, the efficiencies of the rods in th~ plateau
zone may differ in the range of + 30 percent according to deviations of the field
from a uniform plateau in the zone in the range of + 15 percent. Nevertheless,
the experience of operating the RBNIIC reactor shows that in some cases, for exam-
ple, when determininq the effects of yariation of reactivity at which shifting of
a large number of rods occurs the concept of average rod efficiency, which is
equal to approximately 50�10-~', can be introduced, which is less than the reactiv-
ity of the RR rod in the plateau zone since the peripheral rods are also taken in-
to account.
~ ~1 .
~ ' .
a '
~ 1 AP''
~ ~~8 i yCfl~~`
PP
~ 0' 1 '
m 0 6 4?
a (~5
d
~ 0,4 ,
. E ~ij ' '
~ 0,2 �
o~~i.
0 �1 2 ~f 5 6 7�
~2~ rny6uNa noipy~xeHUA Cmep~rHA, M ~
Figure 2.9. Relative Efficiency of Rod
Key:
1. Relative efficiency of rod
2. Depth of rod insertion, meters
The deformation of the neutron field through the reactor radius with shiftinq o~.
the control rods was investigated from the results of the two-dimensional calcula-
tions made. Deformation of the neutron flux near the rod with total and partial
= insertion of it is shown in Figure 2.10. 7.'he degree of deformation was determined
as the ratifl of the mean flux density on a given radius after insertion of the rod
to the flux density prior to insertion of the rod. Averaging was carried out for
all cells located at the same distance from the rod. This ratio is less than 1 for
all rods at a distance up to 12 spacinga of the ahannel lattice from the rod and
the ratio then becomes greater than 1 due to redistribution of the neutron field
thr.o~igh the entire reactar.
Efficienc;~ of rods with film cooling. 7.'he design of control rods with displacers
in the RBMK reactor is not optimum in neutron balance. Actually, a considerable
amount of water that absorbs neutrons remains in it after the rod is removed from
the core. When all the rods are removed from the coref the harmful absorption in
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Table 2.20. Total Efficiency of Rods
~2 Coreonm~e pwxsoQa
c
~T~"R To~~~ . xonon*oe xouop~soe .
~ ~~~~.1 ~
(6) p~ ' ~ p,p443 . p,0309 0~0390 "
( 7) A3 0,0360 0,0247 O,OCi 14
( g) AP 0,0042 0.0021 O,OOGi3
~g~ YCII . 0,0094 0,0074 � 0,0095 ~
-Bce crep~xxx 0,094 0,065 0,088
(10f ~
' Key:
1. Rods 6. RR
2. State of reactor 7. AZ
3. Hot 8� ~
4. Cold with water UsP
5. Cold without water 10. All rods
Table 2.21. Efficiency of SUZ Rpds With Film Coolinq, 10'4
~2i coesoaepe yeaaeopa ~ ~
- . - .
CTlp7KlMb ~~~qtt X0110A80! xQRO~tlO!
C DCOl~O
pp~ 6,s6 5,1 6,47 ,t
. AP~ ~ 3,b6 1.78 2,79
YCIIl8 4,49 3,b4 4.53
Croa6 Bo,(Ig ~ 2,34 3,01 3.06
K,ey :
1. Rods 6. RR and AZ
2. State of reactor ~
3. Hot 8. USP
4. Cold with water 9. Water column.
5. Cold without water ~
the structural elements of the SUZ channels comprises 1.87 percent, including 0.34
percent in the zirconi~am channels, 0.53 percent in the displacers and 1.0 percent
in the cooling water.
The water flowing around the displacer and filling the SUZ channel from the top
and from the bottom makes the qreatest contribution to harmful absorption. The
so-called film cooling system of the SUZ channels in which the flow rate of coolinq
water decreases to ti~e level required to create a surface film approximately 1 mm
thick that cools the pipe of the production channel after removal of the rod from
the core has the best neutron-physical characteristics in this respect. Besides
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~e
u
x 1~0
0
~ qy 2
ai) - ~ ~
_ o o~e
_
- ~
v,~
~
_
-0 �
~ qs ~ ~ ' . ~ .
E � .
0
~ 0,5~ 2 4� 6 8 f0 12 14 l61B 20
( 2 ) PoccmaANae , maP peuremKu
Figure 2.10. Deformation of Neutron Flux Density Pistribution with Variation
o� Position of Control Rod: 1--total ins2rtion of rodt 2--in-
sertion of rod by 2.5 meters
Key : '
1. Neutron flux density, relative units 2. Distance, lattice spacinq
reducing the amount o~ water in the core, this cooling system has a number of
other advantages~ Fox example, the conditions of equalizing the energy release
field through the reactor height are improv~ed (the coefficient of field nonuni-
formity in height comprises approximately 1.2 in the steady mode) and the extent af
fuel burnup increases by approximately 700 MW�day/t. It essentially becomes pos-
sible to control the neutron field by varying the height of the water colu~ in the
SUZ channel. Some characteristics of the RR and AZ rods with film cooling are pre-
sented in Table 2.21. The design and dimensions of the AR and USP rods do not
change and the length of the RR and AZ rods is 6,130 mm. The channels of the RR
and AZ rods operate in the film cooling mode. The efficiency of the rod increases
by 9 percent in the hot state.
The fraction of ab~orbed neutrons in the SUZ channels is presented below with re-
spect to a solid graphite block with different filling of cooling water, 10'4:
Water layer with thickness, mma
0 0.21
1.0 n.30
1.5 0.32
2~~ 0.34
_ 2.5 0.36
3.0 0.38
- Displacer 7 meters long 0.45
water column 2.64
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Harmful neutron absorption in the core with total removal of all RR and AZ rods
with film coo~ing of the channels comprises 0.94 percent, including 0.34 percent
in the zirconium channels, 0.05 percent in the USP displacers and 0.55 percent in
~he cooling water. The average heat release per unit of rod length is 80 W/cm2.
With RR rod lenqth of 510 cm, heat release comprises 41 kW and the maximum specifiC
heat release is 112 W/cm (at Kr = 1.4).
2.5. Nuclear Safety
The core of the RBMK reactor and the fuel assemblies and also the reactor control
and safetiy system and its actuating members ar�e made with regard to the main re-
quirements of nuclear safety regulations of the reactor which is provided in all
operating m~des and states of the reactor and aleo during a~y possible emergency
situations in the production circuit.
The efficiency of 21 AZ rods with minimum rod efficiency of 4.3�10-4 in the cold
- state of the reactor comprises 0.9 percent, which exceeds the value of S in re-
activity with a reserve sufficient with regard to possible "hanqing" of part of the
rods. The efficiency and number of AZ rods was selected on the basis of the maxi-
mum possible rapid variation of reactivity. Variation of reactivity upon "col,
lapse of steam in the core and cooling of the fuel elements to tempexature at in-
= put into the reactor (265�C) and upon dehydration of the fuel channels in a cold
reactor was considered. It was shown by calculations that the values and even the
signs of these effects are considerably dependent on the composition of the core:
on fuel burnup, the number of DP, SUZ rods, displacers and the water columns.
Therefore, states more typical for different phases of reactor operation were se-
lected to determine the effects of reactivity and the nwnber of AZ rods. Of course,
the conditional calculated representation of the core structure was taken into ac-
count in this case. The values of the considered fast-proceeding variations of
reactivity use:d to determine the number of AZ rods are presented in Table 2.22.
Table 2.22. Fast-Occurring Variation of Reactivity as a Function of Extent of
Fuel Burnup
State of Reactor
initial Burnup of Burnup of~ Steady
Effeat of Reactivity ~oad 5 GW�day/t 10 GW�day/t Mode
- Dehydration of fuel channel
in cold state -0.0108 -0.0096 -0.0067 ~ 0.0093
"Collapse" of steam and
"cooling" of fuel elements 0.0083 -O.c~003
According L-o the calculated efficiency of SUZ rods in different states and accord-
ing to the data of Table 2.21, the number of rods should be no less than 17. Tak-
ing into account that the error of calcula~ing the efficiency of a rod may be es-
~timated at + 20 percent, the number of AZ rods was assumed equal to 21. The
minimum efficiency of one group of AR, consis,ting of four rods, comprises 7.2�10'4
and the efficiency of one AR group ~n the hot state comprises 14.4�10-4. With a
rate of displaaement of the rods of 0.4 m/s, the minimum rate of input of reactiv-
ity by the AR rods comprises 0.75�10'5 s'1 = 0.012 S,/s (on the linear section of
the graduated curve?.
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Table 2.~3. Value of ICegf for Different States af Reactor with Charged SUZ Rods
State of Core
Average Extent of Fuel
Burnup, GW�day/t _ workin Cold with water Cold without Water
0(initial charge) 0.913 ~.967 0.963
5 p.g27 0.954 0.950
10 0.937 0.938 0.947
Steady recharging mode 0.917 0.922 0.926
A decrease of the compensating capability of both the SUZ and AP rods due to the
difference of their actual design and arrangement through the core from the ideal-
ized calculated models used was taken into account in the calculations. The re-
sults of calculations for different states of the reactor and different moments
with respect to the run are presented in Table 2.23, from which it is obvious that
the SUZ system provides the required subcriticality of the reactor for all states.
It should be noted that a core structure in which the inserted DP channels were
partially replaced by fresh TVS was arbitrarily assumed for burnup of 10 GW�day/t.
Specifically, the core was assumed to consist of 35 periodicity cells with 2 DP and
two displacers, 20 cells with one DP, one displacer and one SUZ rod and 65 cells
with one DP and two displacers for burnup of 5 GW�day/t; it consisted of 20 cells
without DP and with two displacers, 20 cells without DP and with one displacer and
one SUZ rod and 50 cells with one DP and two dieplacers for burnup of 10 GW�day/t.
The actual state of the reactor will naturally differ from that assumed. There-
- fore, "instructions to ensure nuclear safety during recharging operations in the
RBMK reactor" were worked out to observe the conditions of nuclear safety during
operation of the reactor in the group DP recharging mode.
2.6. Physical and Power Startup of Reactor
Experiments during physical startup of the first unit of the RBNIIC reactor. Start-
up of the RBMK reactor is an important phase of checking the correctness of the
calculated methods and of the applicability of the physical models and finding the
final, most dependable neutron-physical characteristics of the reactor. Special
attention should be devoted to obtaining information required for subsequent op-
eration of the reactor.
The main problems of experiments conducted during physical startup of the reactor
and assembly of full charge of the core reduce to the following. Several ever-
increasing critical systems for estimating the characteristics of the fuel assem-
blies, DP, SUZ rods and graphite stacking are assembled for comparison with exper-
imental data found on a bench or on previous reactors. The minimum critical
system, including the assemblies located in the center of the core, are first
assembled according to the charging cartoqrams the ~ritical systems with DP and
with DP and SUZ rods are then assembled. The reactivity effects when the SUZ chan-
nels, channels with assemblies and with DP are filled with water are analyzed as
- full charge is approached. 7.�he neutron fields are measured in a full charge. The
initial charge of the reactor is shaped with regard to the specific production
characteristics of its components during the final phase by readjustment or removal
of several DP. The results of aalculations are corrected parallel to obtaining the
38
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experimental data. The detailed order of experiments with calculated substantia-
tion for each reactor is determined by the working program of physical startup.
A system without DP with standard SUZ rods removed with dehydrated MPTs circui.ts
and cooling of SUZ reached criticality after charging of 23 fuel assemblies with
two partially charged rods of the ten~porary SUZ).
A system with additional absorbers was then investigated, which made it possi.ble
to determine the critical mass with removed SUZ rods and to estimate the efficiency
of the initially selected composition. A DP composition with ratio of 3:1 of heavy
and light sleeves in the central part 5,000 ~n long and ratio of 1:2 in the end
sections 1,000 mm long each was taken and charging was continued up to 77 periodic-
ity cells from the results of analyzing the experiment and calculation.
Reactivity effects related to filling the SUZ channels and channels with fuel as-
semblies and DP with water were inv~es.tigated in a system consisting of 77 periodic-
ity cells. The experiments were conducted in three phases:
1) study of the reactivity effects with SUZ channels filled with water with
"dry" MPTs circuit;
2) investigating the variation of reactivity with the 1~Ts circuit filled
with water with filled SUZ circuitt
3) determining the effect of dehydration of the SUZ circuit~with MPTs cir-
cuit filled.
Each phase of the investigations was completed by bringing the reactor to the
critical state.
The SUZ channel cooling circuit was filled separately for each group of rods.
Pflsitive reactivity was observed with the channels having charged rods filled with
water, while the water in the channels with remov~ed rods reduced the reactivity
as a result of neutron absorption. The total effect of filling the SUZ circuit
with water was negativ~e and was compensated for by removing 16 standard SUZ rods
and by inserti.ng two RR rods of the temporary SUZ. Filling the MPTs circuit with
water led to determination of po~itive reactivity of 1.75 B. Dehydration of the
SUZ circuit with water in the MPTs circuit was compensated for by introduction of
19 standard SUZ rods.
It was established as a result of the experiments that a system with MPTs circuit
filled with water and dehydrated SUZ circuit has the highest reactivity. However,
further charging was continued in the completely dehydrated core with regard to
the need of conducting production work in the MPTs circuit.
The effect of variation of reactivity with dehydration of channels containinq fuel
assemblies, which was negative and equal to 0.15 6r was determined in this system
of 77 periodicity cells. Remo val of water from DP channels decreased the reac-
tivity of the system by 1.8 R�
39
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~ One of the important integral parts of physical startup was measuring the energy
release fields through the reactor. The purpose of these measurements was as
follows:
. selecting the final arrangement of the fuel assemblies and DP;
analysis of the capabilities of equalizing the energy release field with SUZ
rods ;
substantiation of using the physical calculation pmgrams to predict the en-
ergy release fields;
determination of the effects of the field microstructure and other character-
istics required to process the discrete measurements of the enerqy release =ields
during operation of the reactoz. ~
The measurements were made by small fission ch~mbera whose design was specially
calculated for installation in the carrier tubes of the fuel assemblies with energy
release monitors. The measurements we~e made at eight points through a height of
249 assemblies. The calculated distribution gives a good reflection of the micro-
structure of the energy release field, but yields exaggerated values of energy re-
lease on the reactor periphery. Analysis of the divergence of the experimental
distribution of the energy release field with the calculated value found by the
BOKR-COBZ program showed that the ~tructure of the zone, specifically the extent
of insertion of the rods, the presence of channels under the starting ionization
chambers in the reflector and the axial distribution of the neutron �ield, must be
taken into account in detail for more accurate calculation of the energy release
field. Comparison of the experimental and calculated data found during physical
startup of the reactor of the second unit showed that the mean square deviation of
the calculated d~ta found by the BOKR-COBZ progxam (with regard to the height dis-
tribution of neutrons) comprises 9.7 peraent from the experimental values. This
deviation was also found for calculations made by the QUAM-2 program.
Investigation of the physical characteristics during power startup. Measuring the
reactivity effects and coefficients.
Determination of reactivity effects begins with determination of the integral ef-
fect of variation of reactivity upon heating of the reactor. Calibrated curves
for "weighing" individual rods, obtained in the reactor during heating, are used
to estimate tha variation of reacti~vity compensated for by moving the SUZ rods.
The total temperature coefficient aE that takes into account the simultaneous var-
iation of temperature and water density, fuel and graphite temperature measured
- during first~ heating of the reactor in the temperature range of 100-220�C was neg-
ative and equal to -(4 + 0.5)�10'5OC 1. The total temperature coef�icient deter-
mined upon heating of the reactor after operation for 25 Affective days comp~ised
-(5 + 0.5)�10'5OC-~ in the temperature range of 120-260�C. Comparison of the
states of the reactor with the constants of water temperature and power, but dif-
- fering graphite temperature, permits one to detesmine the temperature coefficient
of the graphite a~. The value of a~ a(3 + 1)�10'SoC'1 upon heatinq for the in-
itial charge of the reactor.
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Precise determinatiAn of the reactivity coefficients, which howewer it is compli-
cated to determine by calculation, ia required to ensure reliable aad safe opera-
tion of the reactor; therefore, a program of experiments to determine the steam
arld power coefficients is realizc,~ in the reactor. An estimate of the steam co-
efficient of reactivity, according to a mode with deviation of one power pump by
45 percent of nominal power, carried out with average fuel burnup of 500 MW�day/t
in the firet unit of the RBMK, yielded a~ m ap/e~ _-0.225� A steam coefficient
of reactivity a~ was detern~ined by deviation of two GNTs upon burnup of 3.5
GW�day/t. The experiment was conducted with reduction of power and low opera-
tional reserve of reactivity (6-8 rods). Processing of the experimental data
showed that the steam void coefficient of reacti.vity became positive: a~ _+~J.7s.
7'he total power coefficient of reactivity (with time constant of the effect less
than 100 seconds), measured at poF.aer of 2,060 MW (t) with burnup of 1 GW�day/t,
was found to be negativ~e: aN m pp/~N =-2.5�10'6 MW"1 (tY~ermal)s aN =-3.2�10-6
MW-1 with burnup of 2 GW�day/t and with power reduction from 1,540 to 1,240 MW (t).
The temperature coefficient of graphite a~ = 4�10'SoC-1 was determined by replacing
tYie helium purging of the graphite atackinq with nitroqen purg~ng with constant re-
actor power and burnup of ~.5 GW�day/t. Processing the graphite temperature by the
readings of standard thermocouples with averaging through the reactor voluime (with-
out regard to the nonuniformity of the graphite block throuqh the voluame) showed
that the increase of the mean qraphite temperature through the reactor per MW of
thermal power is equal to 0.05�C/MW fo~ helium cooling of the stacking and was
equal to 0.1�C/MW for nitrogen cooling. Accordingly, th~ power coefficient of re-
activity through the graphite was equal to 0.2�10-5 MW'1 for helium cooling and to
0.4�10'S MW-1 for nitrogen cooling. The error of detezmininq the data can be es-
- timated at + 30 percent.
Experimental investigation of energy release fields. The distribuLion of the re-
sidual y-activity of the TVS was measured repeatedly in the shutdown reactor dur-
inq operation of the RBMK reactor. The purpose of these experiments was to cali-
brate the radial distribution sensors, to determine the error of discrete
monitoring of the energy release fielrls and to estimate the error of calculating
the energy release fields. Accordinq to the experimental results, the mean aquare
errors of determining the output of the TVS comprise 9 percent forphysical calcu-
lation by the BOKR-COBZ program and 3 percent for Y-scanning.
Satisfactory agreement of the calculated and measured power distribution of the
fuel assemblies made it possible to confidently use the calculated data for oper-
- ational contro~. of the enerqy~release field in the core by using the Prizma pro-
gram in the plant computer. Besi.des the Prizma program, the energy release fields
were monitored and controlled through the reactor by using calculations by a com-
plex of Bazis programs on an external BESM-6 computer. The complex of Bazis pro-
grams combines the program for phyeical calculation of the fielda (BOKR-COBZ), the
_ program of statistical interpolation of channel output from the readings of the
energy release monitoring sensors (Atlas) and the program for thermohydraulic cal-
culation of the channels and calculation of the heat engineerinq reliability
(Zapas). The position of the control members, fuel burnup in the channels, cur-
rents of the intrareactor sensors and distribution of the coolant flow rates
through ~he reactor channels are used as the input data. The power distribution
41
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through all the operating channels of the reactor, the error of determining the
~ output of each channel and t.he reserve to heat transfer crisis in each channel are
determined as a result of calculation. The complex of Bazis progr~ms is a means
df monitoring the operation of the Prizma program.
Recharging the RBMK reactor during operation. A loss of reactivity due to fuel
burnup is restored when the reactor is recharged. Recharging the channels during
operation by means of an unloading-loadinq machine (RZM) is provided in RBr'IIC re-
actors. If putting the RZM into operation is delayed, the channels can be re-
charged in groups (the DP is unloaded during the initial nhase) i.n a shutdown
reactor. The number and location of the removed DP is determined by the specific
operating coniiitions of the reactor prior to shutdown (the reactivity reserve, the
form of radial energy release and so on) and after shutdown tmaximum power and
proposed operating time). Calculated prediction of the next recharging is made by
BOKR-COBZ and QUAM-2 programs, but a depoison~d reactor with dehydrated SUZ cool-
ing circuit (a state with maximum reactivity) is periodically brought to the crit-
ical state af~er each recharging to check that nuclear safety conditions are ful-
filled. The main characteristics of recharging carried out on the RBNffC reactor of
the first unit of the Leningrad AES are presented in Table 2.24 and the change of
the core structure is shown.
All the DP were distributed on several lattices embedded one into the other, which
were recharged in sequence, for convenience in selecting the DP during routine re-
charging. The peripheral DP are exceptions to th3s rule, which is related to the
characteristic features of the SUZ ro@ lattice on the reactor periphery. The DP of
the peripheral group are recharged as necessary with regard to equalization of the
radial energy release field.
- Physical startup of the second unit of the RBMK reactor. Physical startup of the
second unit of the RBMK reactor was begun in May-June 1975. The most important
problem of the experiments carried out during physical startup was comparison of
= the characteristics of the reactors of the first and second units. Systems inves-
tigated in the reactor of the first unit were brought to the critical state several
times for this purpose during assembly of a full charge and a full charge complete-
ly similar to the initial charge of the reactor of the first unit was also assem-
bled. It was established by experiments that al1 the investigated critical systems
have lower reactivity than the corresponding systems of the reactor of the first
unit. This difference comprised 0.5 percent for a fuli charge with water in the
MPTs circuits and SUZ cooling.
Analysis of the calculated and experimental data showed that this divergence can be
exp~ained to a significant deqree by the difference in the average graphite density
in the reactors of both units (1.73 and 1.67 g/cm3, respectively). The effect of
_ dehydration of the fuel channels (0.65 S for the second unit compared to 0.5 B for
the first unit) became somewhat more negative.
The number of DP was reduceci to supplement the deficient reactivity in the reactor:
the initial charge contains 1,455 TVS and 230 DP and eight channels (on the reactor
periphery) remained uncharged. The initial reactor charge is presented in Figure
2.3. Physical startup of the reactor is an important phase of putting an AES into
operation. The startu~i~ a� the RBMK reactors showed that the difference of
42
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~ua~~ad ~(uo~~BZn~ ~ ~ ~ ~ ~
-te~) bu~b.zEq~a~ uodn o M~~~? cv
. . . . . .
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' 43
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characteristics has a significant effect on the properties of the core for tech-
nical reasons of the core components. Therefore, physical startup should precede
introduction of each reactor. The process of physical startup permits formation
of a specific charge of each newly introduced reaGtor with regard to the actual.
characteristics of the core components and to ensure conditions of subsequent YC-
actor operating safety.
BIBLIOGRA.PHY ~
1. Feynberg, S. M., Heterogeneous Methods of Reactor Calculation. Survey of
Results and Comparison to Experiment," in "Materialy Mezhdunarodnoy konferentsii
po mirnomu ispol'zovaniyu atomnoy energii" [Proceedings of an International
Conference on Peaceful Uses of Atomic Ene gy], Geneva, 8-20 August 1955, Vol
5, "Fizika reaktorov" [Reactor Physica], Moscow, Izdatel'stv~o AN SSSR, Z958.
2. Galinin, A. D., "Teoriya yadernykh reaktorov na teplovykh neytronakh" [The
Theory of Nuclear Reactors Based on Thermal Neutrons], Moscow, Atomizdat, 1959.
3. Akimov, I. S., M. Ye. Minashin and V. N. Sharapov, "Develaping Methods of
Physical Calculation of Nuclear Reactors from the World's First AES to the
Present;" ATOMNAYA ENERGIYA, Vol 36, No 6, 1974.
4. Yemel'yanov, I. Ya., M. B. Yegiazarov, V. I. Ryabov et al, "Physical Startup
of the RBMK Reactor of the Second Unit of the Leningrad AES imeni V. I. Lenin," ,
ATOMNAYA ENERGIYA, Vol 40, No 2, 1976.
5. Gorodkov, S. S., "Novyy metod rascheta geterogennykh reaktorov" [A New Method
of Calculating Heterogeneous Reactors], Preprint IAE-2251, Nloscow, 1973.
6. Gorodkov, S. S., "Instruktsiya po pol'zovaniyu pragrammoy rascheta getexogen-
nykh reaY,torov QUAI~iER" [Instructions on the Use of the QUAI~iER Program for
- Calculation of Heterogeneous Reactors), Preprint IAE-2294, Moscow, 1974.
7. Batov, V. V., Yu. I. IG~ryakin, V. I. Pushkarev et al, "The Economics o� the
Transition Period of the Reactors of Nuclear Power Plants," ATONINAYA ENERGIYA,
Vol 26, No 3, 1969.
8. Batov, V. V., Yu. I. Koryakin, V. I. Pushkarev et al, "Selecting the Optimum
Operating Modes of Fuel Charging," in "Nuclear Energy Costs and Economic Devel-
opment, Proceedings of a Sympoaium, Istanbul,"20-24 Oct 1969.
' 9. Yemel'yanov, i. Ya., A. D. Zhurnov, V. V. Pushkarev et al, "Forn?ing the Ini-
tial Charge in a Large Channel-Type Reactor," in "Opyt eksp~.uatatsii AES i
- ~ puti dal'neyahego razvitiya atomnoy energetiki" [The Experience of Operating
Nuclear Power Plants and Methods for Further Development of Nuclear Power
Engineering], Obninsk, FEI, Vol 1, 1974.
10. Palmedo, P. F,, "A Semi-em~irical Description of Detailed Thermal Flux Dis-
tribution," NUCLEAR SCIENCE AND ENGINEERING, Vol 21, 1965.
44
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DESIGN OF A REACTOR PLANT
Moscow KANAL'NYY YADERNYY ENERGE~2CHESKIY REAKTOR in Russian 1980 (signed to press
- 27 Mar 80) pp 48- 79
[Chapter 3 from the book "Channel-Type Nuclear Power Reactor", by Nikolay Antono-
vich Dollezhal' and Ivan Yakov'evich Yemel'yanov, Scientific Research and Design
Institute of Power Engineering, Atomizdat, 2,550 copies, 208 pages]
3.1. The Rsactor
; guse 3.1) with qraphite moderator
~ The channel-type boilinq-water RBMK reactor (Fi
' and water coolant is designed to generate saturated steam at pressure of 70
~ kgf/cm2 (approximately 7 MPa). The main structnral part of the reactor--the core--
i is formed on the basis of calculation and theoretical investigations considered in
~ the previous chapter, where the core structures and its ccmponents are also de-
scribed. The core is located in a concrete shaft measuring 21 X 21 meters and 25
meters deep. The cylindrical graphite stackinq 5 consists of blocks with axial
cylindrical openings assemb~.ed into coluir-.ns in which fuel and special channels are
ins~alled and is located in a sealed cavity ithe reactor apace) formed by the
cylindrical ve~sPl and plates of the upper ar~d lower assembled steel sections.
The reactor space is filled with a mixture of heliwn (approximately 40 percent by
I mass) and nitrogen to prevent oxidation of the graphite and to improve heat trans-
~ fer from the graphite to the fuel channelst leakage of helium is limited by fill-
ing the assembled steel sections and the spaces surrounding the cylindrical ves-
, sel with nitrogen under pressure exceeding the preasure of the helium-nitrogen
I mixture by 20-120 rrm~ Hg (approximately 0.2-1.2 kPa).
i
! The reactor has upper 8, lower 3 and side 4 bioloqical shielding ~.~~ich reduces the
~ radiation intensity durinq operation and all power levels to permissible values
corresponding to sanitary standards im the USSR. The fuel channels (1,693) are
installed in pipe conduit welded into the assem~+led steel sections.
- The coolant is circulated in the reactor circuit by the following scheme. The
coolant--water at temperature of 270�C--is distributed from the pressure vessel of
the ma:in circulating pumps by regulating valves atxd by individual pipelines 2
ttirough the fuel channels. Riaing upward and flowing around the fuel elements,
the water is heated to saturation temperature, is partially evaporated (the aver-
age steam content is approximately 15 percent) ~nd is fed to the separator druins
. 45
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~ ~ ' .
i ; ~ ~
~ ,
~ ,
, , . . . . . ,
- +
. ,y .
,+30600 6 7 8' 10 11
_ ,
, +,~I000 ~ ~ ~ . . ~ +NNO ~ 11
*19320 ~?19200 +193?0
~
. _ 13 ~
- - - - - 14
5
_ ,
S
- 3 , f5
' 00
, p ~
~oo ~ I ' r ~I I? aoo ~
~
J
2 1 ' ' ~ . .
Figure 3.1. Overall View of RBMK Reactor
6 through pipelines 7 in the form of a steam-water mixture. After separation the ~
steam is sent to turbines at a flow rate of 5,400 t/hr at temperatuxe of 284�C and
pressure of 70 kgf/cm2. The condensate from the turbines, passing through t;he re-
qenerative heaters, is mixed with water from the separators and is fed through the
intake water collectors 14 to the main circulating pumps (GTsN) 15, by which it is
fed to the fuel channels.
The nuclear fuel is recharged continuously during operation of the reactor by means
of a loadinq-unloading machine (RZM) 9. The fuel recharging intensity at nominal
reactor power comprises 1-2 fuel assemblies per day during steady operating modet
the maximum RZM productivity is five assemblies per day. The possibility of con-
ducting partial one-time recharqing of assemblies without the RZM on a shutdown
reactor is provided.
The reactor is equipped with fuel monitoring 8ystems which issue information about
its operation as a whole and about the operation of individual fuel channels and
also the necessary signals to the SUZ and emergency signalling syatems
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the physical energy release monitoring syatem through the height and radiue
of the reactors
the system for monitoring the integrity of the fuel channels;
the system for monitorinq the seal of the fuel jackets in each fuel channel
(KGO) 12;
the system for monitoring the water flow rate in the fuel channels;
the system for monitoring the temperature of the graphite and of the assem-
bled steel sections.
The information received from these systems is processed by an automated energy
unit monitaring system.
The main reactor characteristics are presented below:
Reactor power, kW: 3.14�105
thermal 1.106
electrical 37.5�103
Coolant flow rate through reactor, t/hr 5,468
Steam productivity, t/hr ~0
Steam pressure in separator, kgf/cm2
Pressure in group pressure vessels, kgf/cm2 14.5
Averaqe steam content at reactor output; percent
Coolant temperature, �C: 270
at input 284
at output
Maximun channel output with regard to 10 percent 2~g87.6
power mi.salignment, kW 27.95
Coolant flow rate in maximum output channel, t/hr 20.1
Maximum steam content at channel output, percent 1.25
Minimum reserve to critical power 7,000 .
Height of core, mm 11,800
Diameter of core, mm Z50 X 250
Spacing pf fuel lattice, mm 1,693 ~
Number of fuel channela 1,8
Fuel enrichment, percent U-235 1g.5
Average extent of burnup through reactor, GW�day/t ~50
Maximum graphite tempezature at separate points, �C
Maximum surface temperature of zirconium pipe of fuel
325
channel, �C 30
Planned operating life of reactor, years
Assembled steel sections of the reactor. The forces due to the weight of the in-
ternal assemblies, assemblies and piping of the reactor is transmitted to the con-
crete and the inner cavity of the reactor is also sealed bX means of welded as-
sembled steel sections (see Figure 3.1) that at the same time perform the role of
biological shielding. The upper cover 10 serves as the floor of the central room
and at the same time of biological ahielding of the room against the radiation
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of the upper piping of the reactor. The design of the lower part 11 of the cover-
_ ing 10 is made in the form of inetal ducts filled with iron shot and serpentinite.
The graphite stacking is surrounded by water biological shielding located in the
side assembled steel section. The latter is made in the form of a cylindrical tank
of circular cross section with outer diameter of 19 meters and inner diameter o�
16.6 meters. The reservoir is separated inside into 16 vertical airtight compart-
ments filled with water ~hat also remove heat from the g~aphite stacking. The
cooling water is fed to the compartments from below and is removed from the top.
The channels of the starting and operating ionization chambers, drain pipes and
thermocouples sleeves for measuring the water temperature in the compartments are
located in the side structures. The installation space between the outsr surface
of the side structure and the surrounding walls of the concrete shaft is filled
- with sand.
The upper and lower assembled steel sections belong to the more complex and crucial
assemblies. The upper section 8(see Figure 3.1) is a cylindrical shell 17 meters
in diameter and 3 meters high. The bottoms of the shell (the upper and lower slabs)
are welded to it along the periphery by airtight seams and are weided to each
other by vertical stif�ening ribe. Openings are bored in the slabs after rein-
forcing assembly and welding of the structure during installation according to the .
location of the precisely repeated openings in the graphite stacking for the fuel
channels. Channel pipes for the fuel channels and the channels for the control
and monitoring system are installed in the openings and welded and the space inside
the formed reservoir between the pipes is filled with serpentinite. The airtight-
ness of the structure and the quality of welding should meet the reqirements of
helium density. The upper assembled steel section is installed on 16 roller sup-
ports mounted on an annular projection in the upper part of the side steel section
and receives the force from the weight of the charged fuel channels, the slab floor
and the pipelines of the upper piping of the reactor.
The lower assembled steel section 3--the footing for the graphite stacking--has the
shape of a tubular drum 14.5 meters in diameter and 2 meters high. The structure
is loaded by the graphite stacking install~.-1 on it and by the pipelines of the lower
piping of the reactor. Its inner cavity is filled with serpentinite and nitrogen.
The number and arranqement of the lower channel pipes for the fuel channels, welded
into the upper and lower bottoms of the assembled steel section, are the same as in
the upper steel section. After the channel pipes have been welded into the assem-
bled steel section, its inner cavity is tested with a mixture of air and helium to
pressure of 1.25 kgf/cm2 (approximately 0.125 MPa).
The main assembled steel support section 1(see Figure 3.1) in the reactor is
loaded more since it transmits the weight of the lower assembled steel section, the
graphite stacking and the weight of the lower water pipelines to the laying part of
the foundation plate of the building. At the same time its design solution is dis-
tinguished by simplicity and originality. The design is two plates with stiffening
ribs 5.3 meters high perpendicular to each other that intersect through the center
of the reactor. The plat~s are welded alonq the axes of sy~?etry (in the reactor
plane) to the bottom assembled steel section.
Anticorrosion coatings are applied during installation to all the assembled steel
sections of the reactor operating in a gaseous medium with ~he presence of water
vapor.
48
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The assembled steel section of the upper covering 10 in the central room has a
passage for installation of the fuel and special channels. The diameter of the
passage exceeds that of the graphite stacking. The passage is covered by replac-
able floorinq consisting of individual slabs. The flooring plays the role of
- biological shielding of the central room against radiation of the upper piping of
the reactor and the fuel assembl.y when it fa removed from the fuel channel and al-
so serves as heat insulation of the central roo~a.
The slab flooring consists of upper and lower alabs and blocks resting on three
risers. The slabs and blocks of the flooring are assembled stQel sections filled
- with concrete-baritiun-serpentinite cement stone (ZhBSTsK).
_ The space between the upper and lower slabs and blocks of the flooring is used to
lead in the cables of the SUZ servodrives, the energy release monitoring sensors
and thermocouples. Air that then passes through the ventilation duct is pumped
- into the roa~? of the spper piping of the reactor from the central room through
gaps of the plate flooring. The pumped air.cools the plate flooring and elimin-
ates the possibility that radioactive discharges will entar the central room from
the upper piping of the reactor.
Water is delivered and distributed through the reactor channels from group collec-
tors of the lower water piping throuqh preseure-regulating valves and flow meters.
To service, monitor and repair these assembl~es, they are mounted in the passages
below the floor of the room for controlling the requlating valves in the room of
the distributing group collectors. The passages are covered with slabs of con-
crete bi.ological shielding through which rods from the regulating valves are
passed into the upper room.
The graphite stacking. The qraphite stacking (Figuxe 3.2) is assembled on the
lower structure inside the reactor space. It is a vertic~l cylinder assembled
from columns (2,488) consisting of graphite blocks. Each block is a parallele-
piped in shape with cross-section of 250 X 250 mm and 200, 300, 500 and 600 mm
high. The main blocks are 600 mm high while the shortened blocks are installed
only in the upper and lower end reflectors to displace the joints of the blocks
of adjacent columns through the heiqht of the reactor. The overall dimensions of
the core (the graphite moderator) are presented below, the thickness of the end
reflectors is 500 mm and the thickness of the side reflector is an average of
- 1,000 mm. The mass of the stacking is 1,700 tons. Graphite that meets special
requirements in nuclear purity and density is used to manufacture the blocks.
There are openinqs 114 mm in diameter along the axis of the block that form chan-
nels in the columns for the fuel channels and the control and monitoring channels.
Graphite rods 6 are installed in the openings (channels) of the columns of the
si3e reflector instead of the channels.
Each graphite column is installed on a steel support slab 5 which in turn rests on
a support 4 welded to the upper slab of the lower assembled steel section. The
columns are attached and centered in the upper part through sleeve pipes 9 welded
into the upper assembled steel section by means of shieldinq slabs 7 and connecting
pipes 8. The shielding and support slabs are essentially 3dentical in design.
Manufactured of steel, besides performing the~functions of intermediate components
for attaching the columns, they provide thermal shielding of the slabs of the
49
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upper and lower assembled steel sections and are part of the bioloqical shielding
of the reactor.
9 ~
. ~ ~ i0
8 �
7
. 11
6
I
I
I
I .
~
. I
I I I ~ '
1
- 5
4
3 ~ 12
! �
Figure 3.2. Graphite Shielding
A diaph~'agm 2 designed first to create some resistance to the flow of the helium~
nitroqen mi.xture fed through the lower assembled steel section to direct it throuqh
the openings in the support housings into the gap between valves and the stackinq
blocks and second to reduce the heat transfer by radiation from the support slabs
to the upper slab of the lower asaembled steel section, is attached to the supports
by washers 1 and 3. The diaphraqm is made in the form of individual sheets of
08Kh18N10T steel 5 mm thick. ~he gap between the diaphragm and the inner surface
- of the housing of the stacking 11 is clased with a ring 12.
The graphite stacking is secured against displacements in the radial direction by
rods 10 located in the peripheral columns of the side reflector. The rnd is wel-
ded below to the support and is movably connected above to the eleeve pipe welded
into the lower slab of ths upper assembled steel section. The upper joint provides
~ 50
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freedom to temperature shifts of the rod. At the same time the rod is a reflec-
- tor cooling channel. It is manufactured from pipe with outer diameter of 110 mm
and wall thickness of 5 mm. The material is 08IQ~18N10T steel.
Al1 the enumerated assemblies are under inten~ive neutron irradiation and ele~ated
temperatures during reactor operation: thus, for example, the temperature of the
support structures reaches 350�C in the region of the upper lattice of the lower
assembled steel section and 440�C on the lower support slabs and the maximtiun qraph-
ite temperature (calculated) is 750�C.
The temperature conditions of the graphite stacking. The heat from the stacking
is removed to the fuel channels (partially to the SUZ channels), due to which its
temperature mode is determined by heat tra.zsfer from the graphite blocks to the
fuel channels. Gms with an average mass composition of 40 percent helium plus 60
percent nitrogen was used to ensure heat transfer and to maintain the temperature
of the stacking in the range of 700-750�C. Sleeves 8(see Figure 3.3)--solid con-
tact split graphite rings 20 mm high, which were arranged along the height of the
channel tightly against each other such that each alternating ring has direct con-
tact along the lateral surface with either a pipe or with the ;nner surface of a
block and also with each other along the ends, nre placed for this purpose on the
fuel channels. The minimum channel-sleeve and sleeve-block clearances were deter-
mi.ned from the condition of the impermissibility of the channel becoming clogged
in the stacking due to its radiation-thermal shrinkage during operation of the re-
actor. A total maximum reduction of the tolerances on the arder of 1.5-2 mm was
adopted on the baais of operation of uranium-graphite reactors and also data on
irradiation of the reactor graphite, which ensures operation of the reactor for a
long time.
The graphi,te temperature initially increases during operation of the reactor due to
an increase of the end clearances between the qraphite blocks and then decreases
due to the prevailing reduction of the radial clearances. A stable temperature o�
I the stacking is established for approximately 5 years and in this case the temper-
~ ature in the corners of the blocks will be 660�C. The highest temperature of 740�G
(760�C in the corners of the block) is reached duxing the initial period of oper-
ation. The temperature of the outer surface of the fuel channel pipe under the
; solid contact rings does not exceed 325�C. This course of arguments is supported
by data obtained during operation of RBMK reactors at the Leningrad, Kursk and
I Chernobyl'skaya AES.
I 3.2. The Fuel Channel
One ~f. the main assemblies that determines the economy and reliability of reactor
operation is the fuel channel. It is desi.gned for location of TVS with nuclear
fuel and to create a coolant flow. Z'he main heat engineering parameters of the
fuel channel at 100 percent reactor power are pxesented in Section 3.1. The inte-
- gral neutron flux (En > 0.7 MeV) reaches 3�1019 neutrons/cm2 during the calculated
service life of the channel.
The reactor fuel channel is shown in Fiqurs 3.3. The channel housing is a welded
structure whose middle part consists of a pipe 9 with outer diameter of 88 mm and
wall thickness of 4 mm manufactured from Zr + 2.5 percent Nb alloy, while the
51
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upPer 3 and lower 11 end parts welded to it are made of corrosion-resistant pipes
(08Kh18N10T steel) of various diameters. Selection of a zirconium-niobium a?.loy
for the middle part of the channel in the reactor core was determined by the fact
tha~ this alloy has satisfactory mechanical and corrosion properties (a~ > 25
kgf/mm2 (appxoximately 250 I~a), v~,2 ? 17 kgf/mm2 (approximately 170 MPa) and d
is 17 percent) with relatively small thermal neutron absorption cross-section (aa =
_(0.2-0.3)�1029 m2). The middle part of the channel housing is joinec~ to the end
paxts by s~ecially developed steel-zirconium adapters.~
The channel housing in the reactor is arranged in sleeve pipes welded to the upper
2 and lower 10 assembled steel section. It is movably attached in the upper part
by a support fillet and by argon-arc welding of "tendril" seam 4. The lower part
of the housing is connected by welding to the sleeve pipe of the assembled steel sec-
tion through the bellows compensator assembly 12, which permits compensation of
the difference in the temperature expansion of the fuel channel and the assembled
steel sections of the rea~*or and also makes it possible to create reliable air-
tightness of the gas cavity. Moreover, a stuffing-box seal 13 is installed below
the bellows compensator in case the bellows fails. The service life of the channel
housing is calculated at 25-30 years and if necessary it is replaced in shutdown
equipment by means of a special unit which remotely cuts the "tendril" seam between
the sleeve and channel insi~e the upper sleeve and after the channel has been re-
placed also remotely welds this seam and checks the quality by X-ray flaw detec-
- tion. The lowex seam between the bellows compensator and the channel is cut and
. welded by a special automatic welding machine.
The fuel assembly ie installed insid~ the channel on a suspension 5 which holds it
in the reactor core and provides replacement of the spent assembly by means af an
RZC4 without shutting down the reactor.
A plug cap 7 installed in a housing 6 and which seals tlie channel with a gasket is
located on the upper end of the suspension. A solid steel plug 1, which is the
biological shielding is ir.stalled between the assembly and ~he locking plug.
- The steel-zirconium joint. Development of a strong and vacutm~? tight welded steel-
zirconium joint for the fuel chan.nels of the RBNIIC reactor was begun in 1965. Sev-
eral steel-zirconium joints developed and used both in the USSR and abroad were
known by this time. Methods of connecting steel and zirconiian parta by contact-
reactive soldering, explosive welding, joint moldinq and so on were known. How-
ever, all these methods could not b~ used for connecting the pipes of the fuel
channel of the RBMK reactor. The joints produced by these methods were designed
to operate at lower parameters, had lower requiremer~ts on retention of airtightness
and were made with steels having lower level of permissible stresaes at operatinq
temperatures. Therefore, a joint design, the basis of which was the diffusion
welding method, was adopted for the RBMK reactor (Figure 3.4).
Thp inner part of the reducer coupling is made of a zirconium alloy and the outer
part, the enclosing part, is made of stainless austenitic steel. Z'he design of the
adaptor coupling was developed with regard to achieving a configuration and proqram-
med stress state in the joir.t that guarantee strength and re~iability under oper-
ating conditions. The welding technique ensures an optimiun diffusion~interlayer in
52
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. . ; o .o,
� ~ �,o ' 'p.o O
, ,
:o
.o -
b; ~
, ~ n
~ ~ o: o I .
\ . I
.'o I ,
6 �Q: o I 12
_ 5 . ~ .
~a~ 1) B .
~ ~ , . II
i .
~ ,
~
i �
=1 3 . I
~ I , 13 .
2
~ a: o. ~ ~ )
, :o. :o ~ .
::o ~ : .
~ ~ .:o. ~ . . ~ .
_ ' , ;o- ::o~ eoaa ~ ~
I 'o. 'A' I , ' .
I � 10
, . a. :o.
~ :'o:
; :a
Figure 3.3. Fuel Channel
Key : ~
1. Steam-water mixture 2. Water
53
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~ i ~
~95x5
~
095 .
~
~
' ~x~
Figure 3.4. Steel-Zirconium Joint
composition and thickness by which high vacuum tightness and corrosion resistance
in the steam-water mixture and the gaseous medium in contact with graphite are
- achieved. ~The design and technique were developed and the reliability of the
steel-zirconium joint was checked with investigation of the streas state at oper-
ating temperatures, cyclic strength, corrosion reaistance, prolonged corrosive
strength and resistance. Bench and reactor tests of the join con r.ed their
high efficiency during prolonged operation, brief ov~rheating and cyclic thermal
loads.
The adapter coupling is welded to a zirconium allay pipe by electric-arc welding
in a vacuum. The developed design and technique of the welded joint provide
~ strength with high degree of flexibility. Thermal-strain hardening conditions and
machininq of the weld seams and near-seam zones were developed to fulfill high re-
quirements in corrosion resistance. The steel part of the adapter coupling is
- joined to the steel pipe of the channel by argon-arc welding.
The locking device of the fuel channel. The duct of the RBMK fuel channel is
sealed on top by a locking device--plug. With regard to the fact that sealing, un-
sealing and replacement of the fuel assembly operation? should be carried out by an
RZM with remote control, the plug has a simple design that provides reliable con-
ducting of machine operations related to its removal--rotation and vertical motion.
The requirements to ensure airtightness for the entire service life of the fuel
assembly (3-3.5 years) with 30 heat changes during the service life were advanced
- during development ~f the locking device; the sealing gasket should be installed
and remAVed by the same RZM operationst the inner duct should have no sharp turns
and projections to avoid damage to the surface of the fuel element.
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~'he main working members in the plug (Figure 3.5) are a screw and a holder made of
harder steel. During installation of the channel, the RZM w~orks with a special
claw on screw 4, which ensures attactunent of the plug to the channel housing and
consequently ensures sealing. When the TVS is installed into the channel, the eue~
pension housing assembled into a single assembly with the plug, is lowered into the
duct housing. The screw is then tightened upward to the maximum, the balls 8 are
rolled into the bore of the spacer 10 and do not go beyond the outer c~iameter of
the holder. For sealing, the screw is threaded in the nut by the RZM hook and par-
tially forces the balls from the seats of the holder during sealing by the in-
creased diameter of the spacer into the annular groove of the housing. Upon fur-
ther rotation of the screw, the balls, being at rest, prevent longitudinal movement
of the holder upward and create the possibility of grasping the gasket by the clamp-
ing sleeve.
~ llonv~reHUe npvb- -
Ku ne~ed yavv~u~
2epn muaac~uu xa-
~ 1~ . Na~a u np~c eao pcr~-
Pa6ovee no~v- zepNemu3auu~
xeNUe-KaHQ~ (~oKnagKa paayn-
3a~epwema~u- omHe~a) ~ 2 ~
poOaN (npoK~ad-
Ka ynnomHeNa) ~
' 2
3
4 ~
5
~
6
~ ' '8
'9
. 70
i
i 11
' . 12
i �
I . 1~
Fiqure 3.5. Locking Device of Channel: 1--ehank~ 2--flangej 3--semiringt
4--screws 5--locking ring~ 6--channel housingj 7--surfacingj
8--ballj 9--plug housinq; 10--spacert 11--clamping sleevej
12--gasket; 13--suspension housing
Key:
l. Working posit~on-- 2. Position of plug before
channel pressurized beginning of channel
(gasket sealed) pressurization and upon
unsealing of it (gasket
55 unsealed)
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- Installation of energy release monitoring sensors (DKE) through the reactor radius
is provided in 270 fuel channels. The TVS in these fuel channels are distinguished
in design from other TVS by the presence of a sealix~g aleeve from the pipe passing
through the center of the suspension and the fuel assembly designed for installa-
tion of the DKE. When installed in the channel, the DI~ is sealed in the upper
part of the suspension by a metal gasket.
Special channels. Besides �uel channels, special channels in the following quanti-
ties are installed in the reactor:
- SUZ channels 179
Channels with D1~ in height 12
Channels of ionization st~xing fission chambers (I~) 4
Reflector cooling channels 156
Channels outside the fuel lattice to measure graphite temperature: 8
in the platEau zone 4
- in the side reflector
~ in the support and upper shielding slabs 8
Channels outside the fuel lattice for ionization chambers: 20
working 4
starting
The SUZ channels, channels with energy release monitoring sensors in height and
channels of the ionization fission chambers (KD) do not differ from each other.
The design of these channels and their loops is identical and is shown in Figure
- 3.6.
The SUZ, KD and DKE channels (in height) are attached to the upper sleeve pipe by
means of a locking fille.~t and "tendril" weld 3 located on the outside. There are
bellows 2 in the upper sleeves that compensate for the considerable temperature
lengthening of the channels, determined by the temperature drop between the upper
assembled sreel sections and the cold sleev~ pipes. Unlike the fuel channels,
lens compensators 6 are installed on the lower sleeve pipes of the special channels.
The upper and lower parts of the special channels are made of corrosion-resistant
steel, while the middle part is made of zirconium-niobi alloy. The middle part
is joined to the upper and l~wer parts of the ch nel by steel-zirconium adapters
similar to those of the fuel channels.
The SUZ channels 1 have heads 4 designed to attach the actuating mechanisms and to
deliver cooling water to the channel. The SUZ actuating mechanisms 5 are attached
to the heads of the SU7, channels, while sealing sleeves are attached to the heads
of the DKE and KD channels, also on gaskets. The sleeves in the DI~ channels are
designed for installation of sensors and are made of SAV-1 aluminum alloy and the
sleeves in the KD channels are designed for installation of the suspensions of the
ionization fission chambers and are made of corrosion-resistant stePl. A permanent
throttle valve 7 whose designation is to create resistance to water flow through
the channel that provides reliable filling of it with water, is installed in the
lower part of the special SUZ, DKE and I~ channels.
The reflector cooling channel (Figure 3.7) is designed for cooling the side re-
flector of stacking 4, the upper asse~nbled steel section of the rods for attaching
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5 ~
4
3 ~ Bo9a I
(1)
~
2 ~
�
6
~ ~M
. k'
I
I � be
i �?z~ . ~
1 I ' 1
(
I -
. . . , ~ � (l~
~ . ' ~ . . ~'BoA~
Figure 3.6. SUZ Channel ~
Key:
1. water
the side reflector 5 and also to reduce the thermal flux to the housing and compen-
sators which form the inner airtight cavity of the reactor. The channel is struc-
turally made in the form of a Field pipe of corroaion-resistant steel. Water enters
the channel from the top through the central pipe and is removed through the space
between pipes, rising upward.
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,
. ~ .
-
'.n: . c v . . .
. .v I .
~
.a��~�p I .
n: ~a' � /
~ 1 ~ Bada 4
f-
5 I
I
. I
6
3
_ 7
2 I
~ 8 ~ .
.
'
~ : ;
. ~ . . . .
Figure 3.7. Reflector Cooling Channel: 1--lcop sleevej 2--upper assembled
steel section= 3--bellows compensator; 4--side reflectorj 5--
rod for attaching side reflectort 6--Field pipe= 7--sleeve;
8--lower assembled steel section
Key :
1. Water
3.3. The Reactor Pipelines
The reactor pipelines provide circulation of coolant in the channels and assembled
steel sections and also circulation of gaseous media in the reactor and the reactor
space. The distributing lines contain pipelines and collectors, their suspensions
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and supports, fittings with drives and production monitoring devices. The pipe-
lines are divided into lower and upper accordinq to conditions of installation and
location. The first group includes water pipeli.nes to deliver water to the fuel
- channels and drain pipelines of the SUZ channels. The second group includes
water-steam pipelines that remove the steam-water mixture from the reactor, pipe~
lines that deliver water to the SUZ channels, pipelines that deliver and drain wat-
er of the reflector cooling channels and also pulsed pipes of the system for mon-
itoring the integrity of the fuel channels (KTsTK).
The water and steam-water distributing pipes are part of the multiple forced cir-
culation (I~Ts) loop in which the coolant is transported in the following.manner:
water from the two delivery collectors of the main circulating pumps (GTsN) enters
44 group collectors (22 each on each side of the reactor). The water is delivered
from the group collectors through a locking-control valve and spherical flow meter
through an individual pipeline to each fuel channel. The coolant-steam-water mix-
ture--is fed through the upper sleeve in the fuel channel into the pipeline
through which it is fed directly into a separator. The diameter of the water
distributing pipelines is equal to 57 mm and the wall thickness is 3.5 mm. Guide,
movable and fixed supports and suspensions are provided to ensure the efficiency
of the pipelines. The pipelines of the steam-water distributing pipes 76 ~n in
diameter and with wall thickness of 4 mm are separated alternately on both sides
~ of the reactor, symmetrically with respect to the axial plane. The rows are ar-
i ranged in layout within the upper assembled steel section with spacinc~ of 250 mm
and are arranged in the separator room with variable spacing of 250, 500 and 1,000
~ mm. The corresponding connecting pipes of the separators are arranged along the
length of the separator with spacing of 250 mm.
I Water is fed to the SUZ channels and the reflector cooling channels thraugh indi-
j vidual pipelines from a common delivery collector. The pipelines of the SUZ chan-
nels are separated in bundles along the corresponding rows of the sleeves and are
- connected to the channel heads by tendril welding. Water is removed from the SUZ
channels through pipelines located under the lower assembled steel section into a
drain collector. Locking-regulating valves and flow meters are installed on each
pipeline that delivers water to the SUZ, Dl~.and I~ channels. The valves have
man~ual control. An individual throttle washer or locking-control fitting that
serves to distribute the flow rate ~hrough the channels during adjustment is in-
~ stalled at the inlet to the housing of each flow meter of the reflector cooling
channels.
-i The impulse pipes af tt~e KTsTK system are designed to remove the gaseous mixture
' from the reactor space in the region of each cell to check the tightness of the
channel. The helium-nitrogen mix ture is pumped out to the purification system
through the impulse pipes during normal operation. Detection of water in the gas-
eous mixture being pumped out indicates a leak from the fuel channel.
Besides the considered distributing ~ipelines, there are additional pipe systems
on the bottom and top of the reactor, for example, pipelines for delivery and re-
moval of nitrogen to the dif~erent zones of the reactor and the assembled steel
sections, for delivery of the helium-nitroqen mixture (discharge of the steam-
gaseous mixture), for delivery and removal of water from the side assembled steel .
sections and different drain pipes.
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3.4. Flow Regulators
All uranium-graphite reactors have devices which are used to maintain the flow
rate of the working medi ~n at a specific level or to regulate the flow rate in the
required range. These devices, namely the shutoff-regulating valves, are installed
in the first loop of the RBMK reactor, ar the inlet to each fuel channel, and are
designed to regulate the coolant flow to achi.eve a specific steam content. The
fluid flow ra~e through the object of regulation is usuall.y varied by the throttle
valve method in nuclear and power plants. ,
The shutoff-regulating valve (Figure 3.8) provides the necessary regulation and
possibility of preliminary monitoring of the water flow rate through the fuel chan-
nel in all operating modes of the reactor and also cutoff of the fuel channel from
the group collector during repair of the channel or pipes of the water distributing
lines. The valves are installed in the room of the water distributing lines on the
group collectors and are connected by rods to the indicators and control levers
located behind the concrete cover. Water enters the valve cavity from the group
collector, passes through the throttle valve device and the flow meter and pipe of
the water distributing line to the fuel channel. Regulation is accomplished by
changing the gap between the end piece and the seat of the throttle valve. The
valve should provide continuous reliable operation for 50,000 hours (Table 3.1).
Table 3.1. Characteristics of Regulating Valve
IInit of Calculated Experimental
_ Parameter Measure Value Value
Flow rate with full-scale parameters t/hr 30.6-12.0 25.4-12.3
- m3/hr 39.7-15.6 39.2-16.0
Pressure drop kqf/cm2 6.2-13.9 7.6-14.0
Travel of shutoff inember mm 9.6- 3.6 7.7- 3.7
Area of narrow cross-section of seat cm2 2.5- 0.9 2.0- 0.9
Throttling gap mm 2.3- 0.9 1.9- 0.9
Coolant flow rate m/s 44.1-48.1 47.5-50.4
Dynamic head in narrow cross section kgf/cm2 7.7- 9.0 8.3-10.1
Reserve to boiling kgf/cm2 18.0- 9.0 15.8- 7.6
Maximum cold (hot) water flow rate m/s 70(80) 70(80)
3.5. Selecting the Structural Materials and the Water Chemical Regime
The main construction materials u~ed in an AES with RBMK reactor are stainless and
perlite steels, zirconium alloys and NR~iZh~S-1 copper-nickel alloy. Stainless steel
08IQ118N10T is one of the main construction materials of the MFTs loop (pipelines,
plating of the separator housing and part of the fuel channel). The total stain-
less steel surface in the MPTs loop comprises approximately 25,000 m2. Steel of
this type is used to manufacture the housings and pipe bundles of the low-pressure
heaters.
Perlite steels are used for the pipelines of the condensate-feed channel, the con-
denser housings, the heating steam pipelines and the saturated steam pipelines. The
total surface of perlite steels comprisea approximately 5,000 m2. Perlite steel is
used in the MPTs loop for large shutoff fittings.
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. ~ .
. 4
2 5
~ 6
7 '
d
-
'�0: ' A: .
� 9
p .':p: . :Q:
f0
.:p; ~1l '
::o: ..~:,..o,: n
~ . .~a : .
~ - ~
~ .
I ,
;
~ . .
3
I ' ~ '
. , .
.I ~ ~ -
� ~s
\ . .
. ~ .
~ . - .
~ ~ ~ .
Figure 3.8. Shutoff-Regulating Valve: 1--indicator; 2--indicator screwi
3--drivet 4--drive shankj 5--threaded sleevef 6--connectori
7--ringj 8--shafti 9--sleeves 10--gasket; 11--bellows; 12--
housing~ 13--throttle valvej 14--end piecef 15--throttle valve
seat= 16--group collector ~
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The fuel channels, SUZ channels and fuel element jackets are made of zirconium al-
loys. The total surface area of zirconium alloys is 13,500 m2. The pipe bundles
of the turbine condensors are made of copper alloys. '
The distinguishing feature of the water reqime of a boiling-water reactor is in-
creased oxygen concentration in the coolant, caused by radiolysis of water in thp
core. Consequently, the materials for the MPTs loog are selected with regard to
this circumstance.
- Neutron irradiation essentially has no effect an corrosion of 08IQz18N10T steel.
Steel corrosion is not intensified under shutdown conditions. It may be subjected
to corrosion cracking with~the simultaneous presence of chlorine ions and oxygen
or another.oxidant in the medium and with the presence of tensile stresses in the
- metal. Cracking was not observed in stressed metal in a mediiun containing only one
of these agents [1]. There were no cases of corrosion cracking of the metal (at
Q= 33 kgf/mm2 (330 MPa) during 10,000 hours of testing) in water containing 0.1-
0.3 mg/kg of oxygen and 0.1 mg/kg of the chlorine ion at 285�C and in saturated
steam containing up to 20-40 mg/kg of oxygen.
Corrosion cracking of OS1Qz18N10T steel is observed with chloride concentration on
the surface. Concentration of the chlorine ion on the surface of steel in satur-
ated steam containing up to 30-40 mg/kq of oxygen, at which corrosion cracking of
steel is observed after 6,000 hours, comprises 1�10'3 mg/cm2. There may be a phase
interface of the med'ium in the separator at the possible location of chlorine ion
concentration under operating conditions of an AES with the RBMK reactor. However,
compressive stresses are prevalent at the interface.
An estimate of the operating life of stainless austenitic steel under the operating
conditions of a boiling-water reactor, carried out according to the func~ions from
~ [2], indicate that there should be no danger of corrosion cracking in the MPTs Zoop
during 25-30 years of operation if normal water regime is observed. The danger of
corrosion cracking determined by chlorine-ion concentration on the outer surface of
MPTs pipelines due to possible leaks from the outside is prevented by electrochem-
ical shielding made in the form of an aluminum coating with AS-8A organosilicate
material.
The assembled steel sections of the reactor of perlite steel are made with the same
coating. Extensive use of perlite steels both in the MPTs loop and in the condens-
ate-feed channel is still limited since additional purification of the coolant is
required, especially after shutdown modes. The feasibility of replacing stainless
steel by perlite steel can be solved from the results of prolonged operation.
The high corrosion resistance of the zirconium-niobium alloys used in RBMK reac-
tors in water and steam at 285�C is explained by their capability of passivation
as a result of protective oxide films forming [3]. Zr + 1 percent Nb alloy is used
for the fuel jackets and ZR + 2.5 percent Nb is used for the fuel channels. The
oxygen contai.ned in the water affects these alloys at 285�C.
Reactor radiation is a specific factor whose effect should be taken a.nto accourct
when estimating the resistance of zirconium alloys. It was established experimen-
tally and confirmed theoretically that the rate of the anode process and
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consequently the rate of corrosion of Zr + 2.5 percent Nb alloy increases by only
5-10 percent under reactor irradiation conditions. The rate of corrosion of Zr +
+ 2.5 percent Nb alloy does not exceed 0.024 g/(m2�day) during 8~000 hours of
testing. The intensity of removal of corrosion products to the coolsnt comprisea
1-5 percent of the corrosion rate. The efficiency of zirconium-niobium alloys in
the components of the fuel channel, welded joints and adapter is determined by pro-
longed laboratory and industrial tests I4-10].
Stable nperation of RBMK reactors confirr.:s the correctness of selecting materials
and corrosion rates of structural materials for the MPTs loop. Taking the selec-
tion of construction materials in an AES with RBMK reactor and also the character-
- istic features of the corrosion behavior of these materials into account, a cor-
- rectionless water regime was selected and calculated with provision of high purity
in the main water loops without any correcting additions and the minimum possible
oxygen content in the coolant.
It should be noted that steady oxygen concentration due to removal of radiolysis
products and the difficulty of recombination processes has not been established
during radiolysis of water in boiling-water nuclear reactors of the RBMK type op-
erating in a single-loop AES circuit. The final products of water, hydrogen and
oxygen radiolysis are removed by the steam-water mixture to a separator. They are
delivered together with steam from the separator to the turbine condenser where
they are removed from the loop.
The radiation-chemical reactions of water in the reactor determines the stochia-
metric yield of hydrogen and oxygen of approximately 0.6 molecules/100 eV and 0.3
molecules/100 eV, respectively. In this case the oxyqen concentration in the sat-
urated steam of the separator comprises approximately 10 mq/kg and that in the
water of the MPTs loop during steady operating modes comprises 0.03-0.05 mg/kg.
Suppression of radiolysis by special hydrogen additives or hydrogen-containing
compounds for boiling-water reactors operating in the single-loop AES circuit is
nnt required.
Table 3.2. Design Standards of Coolant Quality of AES with RBI~IIC Reactor
Feed Water
~ Water of After Conc3ensate
Indicator MPTs Circuit Purification
pH value 6.5-7.2 7+0.2
Specific electric conductivity, ucm/cm 1.0 0.1
Concentration, mg/kg:
of chlorine ions 0.1* 0.004
of iron** 0.2 0.01
of copper 0.05 0.002
of oxygen Not regulat$d 0.05
salts of hardness, ug-eqv/kg 15 0.5
*
A Concentration of 0.15 mg/kg is permitted briefly up to one day. .
Up to 1 mg/kg during transient and starting modes.
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If the quality of the reactAr water is normalized, it had the purpose of providing
minimum rate of corrosion o� theaco ssfo �~enq~lityaof the feed water were deter-
on the fuel elements. The stand rd eva ration of the reac
mined with reg$rd to the coefficiEnt of concentration by P~
~or water, the corrosion rate of the ~~er~ion.~f~eesta dardstadopt dcfornthe~a
= the possibility of the condensate purifica
- design stage are presented in Table 3.2 for the RBMK reactor.
The standards are refined and correc~ed during operation. Thus, upon conversion
to automatic manitoring of pH, the normal range of this indicator was expanded to
6.5-8.0 for reactor water and the siandards az'e ~aee
rseofloperating AES withlthe
products of iron, copper and hardness salts. ManY Y
RBMK reactor showed that the real w~.ter regime is considerably better than the
planned regime. The indicators of the water of the MPTs loop are presented below:
0.4-0.6
. Specific electric ~onductivity, uCn?/cm
Content, mg/kg: 6 ~
C f5 ,
~ ~
- ~
. ~
13 ~ . ~
12
17
, 20
B ~ 18 19 . .
10
i -
-I '
_i ' ~ 9
i ~
- 6
~ ~ 5 2~
4
. ~ ~
3 2 9 22
Figure 10.3. Functional Diagram of Drives for Recharging and Unsealing of
RZM: 1--fuel channel (TK)~ 2--shutoff plug of TVS suspension~
_ 3--spherical lock of plugj 4--joining connection pipe (SO);
5--grab; 6--TK sealing keyj 7--moment clutches of TK sealin~
drive; 8--TK sealing drive; 9--damping spring of gram; 10--
spherical nuts of TK moving drives; 11--spherical screws that
limit the rate o~ motion of the TKt 12--chains that impart mo-
tion= 13--reducer with moment couplingst 14--motor; 15--car-
- riage with sprockets; 16--flexible elements of strain gaugesj
17--switch to raise SP 18--bellows seal of SP; 19--moment trans-
�er clutch~ 20--TK moving drive~ 21--sealing cuffs; 2--concrete .
blocks
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The following machine service zones are provided in the central room.
1. The parking place is a zone in the central room designed for parking the ma-
chine during periods between reactor recharges.
2. A simulator bench designed for:
adjustment and checking of the mechanisms o.f the machine;
filling the pressure vehicle with condensate;
simulation of standard recharging~
loading a fresh assembly into the pressure vehiclet
decontamination of the inner cavity of the pressure vehiclef
~ replacement of the inflatable cuffs of the j~ining connection pipe.
The simulator bench has the corresponding equipment to perform.these operations.
3. An assembly for receiving the spent assemblies is used to receive the gauge.
4. The repair zone is designed to replace a pressure vehicle that has failed. It
is located in the central room in the region of the simulator bencY:. A fully as-
sembled reserve gressure vehicle is permanently located in the zone.
10.2. Operating Modes
_ Preparation of the RZM for operation. The machine is put into operation in the
following manner. The simulator bench is prepared: the plugs are removed �rom
the seat (seats I and II of the simulator bench) to fill the pressure vehicle with
- condensate. The gauge and production plug are installed in these seats. The
_ electropneumatic power supply of zhe machine mechanisms is switched on and it is
fed from the parking place to the simulator ber.r.h i;o seat I.
After power has been supplied to the RZM, the valve for filling the feeder tank
" with condensate from the loop is opNne~,,. automatically. The machine is joined to
seat I on the simulator bench and the ga~.ige is tight;,~ned into the dry machine and
, the machine is then joined to seat II and the production plug is removed from it.
The pressure vehicle of the RZM is then filled with condensate in seat II until
the level in the filling tank of the joining connection pipe appears. After the
pressure vehicle has been filled with condensate, the sealed slide valves are
opened, the condensate is dumped from the seat and the joining connection pipe to
the pipi: g in the machine is separated from the seat.
. Standard recharging. The fe~llowing procedure is observed when the machine is op-
erating in the standard recharging mode. The RZM with pressure vehicle filled
with condensate (t = 30�C) is directed toward seat I or II of the simulator bench
in which fresh fuel assemblies are first installed. The joining connection pipe
is joined and sealed with the head of the seat. The seat and joining connection
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pipe are then filled with condensate from the seat filling system. 'I'he slide
valves are then opened after this operation, a fresh TVS is screwed into the car-
tridge of the pressure vehicle and is installed in it. The protective slide valves
are closed and the condensate from the joining connection pipe in the seat are
, dumped into the sewer.
The machine is disconnected from the seat and is sent to the reactor to recharge
the fuel channel.
Recharging of the fuel assembly. The operator instructs that a specific fuel chan-
nel be prepared for recharging according to the program for recharging the reactor.
The protective plug is removed from the channel and the cable of the Y--detector is
disconnected from the assembly plug. These operations are performed ma.~ually by
workers of the central room. The RZM automatically emerges to the coordinate of
the channel to be recharged, is joined and the plug of the machine is then sealed
- to the head of the channel by high pressure. The joining connection pipe is then
filled with condensate from the tank, the slide valves are opened, the feed pump
is switched on and a pressure of 73-75 kgf/em2 is created in the pressure vehicle. ~
- The grab is lowered and linked to the head of the assembly. The sealing and un-
sealing mechanism is switched on and the channel is unsealed.
When the process of unsealing is completed, the hoisting mechanism is switched off
and the spent assembly is raised to a heigh~ of 7.5 meters to the cooling zone
where it is held for 10 minutes. The cold condensate (t = 30�C) begins to be fed
into the fuel channel at the moment it is unsealed from the machine (the pump op-
~ erates to feed condensate constantly to the pressure vehicle until a fresh assembly
i has been installed in the channel and it has been sealed) at a flow rate up to
-i 0.5-1.0 m3/hr. After the assembly has been held in the coo?ing zone, the hoisting
mechanism is switched off, the assembly is inserted into the pressure vehicle and
is installed in the cartridge. The passage of the channel is checked by the gauge
and a fresh assembly is then installed in the reactor. The pump feeding the con-
densate to the pressur.e vehi~le is switched off after the channel has been sealed
and the pressure in the pressure vehicle is dropped to atmospheric pressure. The
sealing slide valves are closed, the cavity of the connection pipe is joined to a
- special blower and the airtightness of the channel head is checked. The condensate
is removed by compressed air delivered to the upper part of the joining connection
pipe, which forces the condensate into the tank, prior to disconnecting the connec-
tion pipe with the channel from th~ cavity of the joining connection pipe.
I
-i
~ The RZM is then unsealed and disconnected from the fuel channel and the machine is
sent to the assembly for reception of spent assemblies. The receiving assembly is
prepared for unloading the spent assembly from the RZM pressure vehicle into the
jacket of the holding tank, which is installed in one of the seats of the receiv-
ing assembly, even during operation of the RZM on the reactor. The jacket is
filled with condensate. The machine with the spent assembly in the pressure ve-
hicle automatically emerges to the coordinate of the prepared jacket. The machine
is joined and sealed to the jacket of the holding ~ank. The cavity of the joining
connecting pipe is filled with condensate from the tank, the slide v~lves are
opened and the spent TVS is removed from the pressure vehicle to the jacket. The
slide valves are then closed, the condensate is expelled into the tank from the
connection pipe by compressed air and the machine is disconnected from the jacket
and is ready for a new recharging cycle.
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Recharging a shut-down and cooled reactor. A shut-down and cooled reactor can be
recharged by two versions:
1) unloading two spent assemblies with installation of fresh assemblies to
replace them=
2) unloading two spent assemblies and fresh assemblies are installed in the
- reactnr v~ithout using the RZM by means of devices and transport equipment pro-
vided in the central room.
In both versions the production operations of the RZM are simplified in nature
since the excess pressure in the reactor is mai.ntained at 2-5 kgf/cm2 (approxi-
mately 0.2-0.5 MPa) or is absent and the condensate is drained to the level of
the assembly heads. If thEre is pressure in the reactor~ the RZM is continuously
filled with condensate and it is not delivered to the pressure vehicle by the RZM
systems. If there is no pressure in the reactor, condensate must be fed to the.
pressure vehicle constantly when unloading the spent assemblies. The production
plug is first removed from the cartridqe of the pressure vehicle when unloading two
_ spent assemblies and the gauge is also removed when unloading four assemblies. The
time of the recharging cycle for two assemblies is 350 minutes and consequently the
RZM is capable of recharging eight fuel channels per day. A total of 267 minutes
is expended to unload four asse:+hlies, i.e., the RZM is capable of unloading 20
fuel channels per day.
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PROSPECTS FOR DEVELOPI~NT OF LTRANIUM-GRAPHITE CHANNEL-TYPE REACTORS
Moscow KANAL'NYY YADERNYY ENERGETICHESKIY REAKTOR in Russi,tn 1980 (signed to press
27 Mar 80) pp 189-203
(Chapter 11 from the book "Channel-Type Nuclear Power Reactor", by Nikolay Antono-
- vich Dollezhal' and Ivan Yakovlevich Yemel'yanov, Scientific Research and Design
Institute of Power Engineering, Atomizdat, 2,550 copies, 208 pages]
11.1. Principles for Improving the Core
Channel-type uranium-graphite power reactors will be improved in their subsequent
~ development in engineering and economic characteristics. These improvements will
i be made as further design and planning developments are made, experimental inves-
tigations are conducted and experience in operating reactors at existing AES is
accumulated. As a result capital expenditures per unit of established gower of an
AES, the cost of generated energy, the number of operating personnel, specific con-
struction volumes and consumption of inetal, including stainless steel, the labori-
ousness and periods of constructian and installation work should be reduced and
the operating reliability and maneuverability of the enerqy units with these reac-
tors should be increased. All this can be achieved by increasing the intensity of
the core and the extent of nuclear fuel burnup, conversion to the block principle
of constructing a reactor, i.ntroduction of nuclear superheating of steam with an
increase of its initial parameters ar,d with simultaneous improvement of the thermal
layout of the unit, improvement of the configuration of the energy block and other
- improveme~nts in the designs of the basic equipment and of schematic and configura-
tion decisions.
One must also bear in m3nd in this case the tendency toward an increase of the
u~^it output of units caused by the constant increase of the requirement of the
I total introduced power of AF.S. Some promising solutions to improve uranium-
graphite power reactors and energy units with these reactors are considered below.
An increase of the unit power is one of the possible methods of improving the
economy of AES. This path was reflected in development of RBNIIC-1500, RBMK-2Q00
and RBMKP-2400 [1-4] with electrical output of 1,500, 2,000 and 2,400 MW, respec-
tively, with respect to RBMK reactors. The output of the fuel channel has baen
increased 1.5-fold in the RBMK-1500 reactor compared to the RBMK-1000 reactor due
to intensification of the heat transfer, which makes it possible to increase the
power of the BRMK-1000 to 1,5000 MW while maintaining the overall dimensions and
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design of the reactor. The chanx~el diameter, number of fuel elements i.n the as-
sembly and lattice spacing have been increased in the RBNIIC-2000 reactor. This
makes a.t possible to develop a reactor with double output in the diunensions of
the RBMK-1000.
Development of a channel with increased output with respect to RBNIIC reactors opens
new opportunities toward improving their engineering and economic characteristics.
The experience of operating AES with RBMK-1000 reactors showed that there is some
reserve both with respect to the linear loads on the fuel elements and with respect
to the maximum channel output (with respect to conditions of heat transfer crisis).
On the other hand, it turned out that the stability of energy release distribution
decreases as fuel is burned up and as additional absorbers that compensate for the
initial excess reactivity are removed from the reactar.
- As indicated by calculations and confirmed by experiments, the main factor that
determines the deformation of the energy release fields with time constant of the
first azimuth harmonic from several to tens of minutes is a positive steam reac-
tivity coefficient a~. A decrease of its value improves the stability of energy
release distribution. An increase of the nuclear fuel to moderator nuclei ratio
can be regarded as the most economical and optimum method of reducing the steam
coefficientj an increase of enrichment is most economical and optimum for operating
reactors. The main characteristics of a reactor with an increase of fuel enrich- ~
ment for steady continuous recharging mode are presented in Table 11.1.
Conversion to increased fuel enrichment leads to an increase of the extent of burn- '
up, variation of coefficient a~ that improves field stability and to a reduction of
nuclear fuel and fuel element consumpt~on. On the other hand, an increase of en-
richment and a corresponding increase of the extent of fuel burnup lead.s to an in-
crease of output of a freshly charged channel, to an increase of the lifetime of
the fuel in the reactor and to an increase of linear loads on the fuel element
(Table 11.2). The possibility of increasing channel outFut (up to 1.5-fold) has
been confirmed by development and creation of the fuel channel for the RBMK-1500
reactor.
Turning to discussion of the results, it should be noted that the given data
should be regarded mainly as comparative rather than as absolute data since the
calculations were made for a steady fuel recharging mode. ioefficients of nonun-
iformity Kr and KZ, found with regard to the experience of operating RBMK-1000
reactors, and the calculated recharging coeffic3.ents were used when determining
- the output of freshly charged channels. Kr = 1.4 was adopted for reactors operat-
ing on fuel with enrichment of 1.8 percent and t}xe coefficient of nonuniformity
by height was Kz = 1.4. Based on the experience of operating RBt~IIC-1000 reactors,
_ the mean square error of determination and maintenance of channel output Qk was
_ assumed equal to 5.2 percent while that of the linear load on that of the fuel
element was Qt = 7.7 percent.
Z'he data indicate that an increase of fuel enrichment in RBMK-1000 reactors is a
realistic method of increasing the effectiveness of fuel utilization. The use of
enrichment up to 3.6 percent is possible according to the permissible outputs of
channels and linear loacls on the fuel element. The problem of increasing channel
output and the linear loads on the fuel element up to 1.5-fold has been resolved
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ro
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. 115
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116
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in the RBMK-1500 reactor [5, 6~. In this case the TVS are equipped with heat
t~ansfer intensifierss the channel design and the delivery and drain pipelines are
not ahanged. TVS without intensifiers can be used with more moderate increase of
fuel enrichment in the RBMK-1000. For example, conversion to 2 percent fuel en-
richment can be accomplished without any design changes in the TVS.
We also note that the maximum values of output presented in Table 11.2 were ob-
tained on the assumption that the algorithm for equalization of output through the
cor~ remains the same as in existing RBMK-1000 reactors. At the same time one can
suggest a number of ineasures, part of which has already been tested experimentally
in reactors, that make it possible to expand the capability of equalizing the en-
ergy distribution in the core. These measures may include the use of absorbing
~ rods inserted into fresh TVS and removed as channel output decreases, the use of
- burnup absorbers, optimization of recharging procedures and so on. A specific
reserve for equalization is related to the possible increase of the operational
_ reserve of r~activity with some operating modes of the AES in the power system.
An increase of enrichment leads to a decrease of fuel element and natural uranium
consumption. The annual consumption of fuel elements upon convezsion from version
to version decreases by 20-30 percent and if three percent enrichment is used, it
decreases by one-halfs the U-235 content in the unloaded fuel is reduced to 0.25
percent, i.e., to the spent tailings of enrichment plants, and if enrichment of
3.6 percent is used, it is reduced to 0.2 percent, which resolves the question of
the feasibility of extracting the U-235 from the spent fuel. An increase of in-
; itial enrichment significantly changes t�he steam coefficient of reactivity, re-
ducing it compared to enrichment of 1.8 percent and even shifting it in the nega-
i tive direction, 4~hich proves the stability of energy release distribution, but
requires specia] consideration of the transient modes.
~ If there is no doubt of the variation of parameters of channel output, linear load
on the fuel element and variation of the steam coefficient of reactivity a~ ccn-
siderad above with an increase of enrichment, an increase of the extent of burnup
to 40-46 GW�day/t r~f uranium and the calendar lifetime of the fuel elements in the
core to 10-12 years when using enrichment of 3-3.6 percent accordingly requires
special investigations and confirmation. The prognosis for development of fue3
elements with oxide fuel burnup of 45-50 GW�day/t of uranium may be assumed real-
istic on the example of fuel elements for WER reactors [7, 8]. A more difficult
problem is to provide viability a.nd integrity of the fuel elements for prolonged
periods under reactor conditions.
~ The possibility of increasing channel output 1.5-fold opens up new prospects for
~ RBMK-1000 reactors and ensures a significant improvement of the economic indicat-
ors of the fuel cycle due to an increase of fuel enrichment. When fuel with en-
richment of 2.4-3.0 percent is used in RBMK-1000 reactors, development and
experimental confirmation of the efficiency of fuel elements that permit extent
of burnup up to 40 GW�day/t of uranium and that have a guaranteed lifetime of
approximately 10 years in the core are required. This is one of the trends toward
which extensive experimental.-design and research work should be condL~cted to im-
prove RBMK-1000 reactors [11] .
The outlined concepts are based on the assumption that the fuel is uranium dioxide.
If favorable results in development of the techniques of other fuel compositions~
~ 117
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for example, silicided, carbide and other fuels, they can be usedj calculations
indicate their promise.
ii.2. Sectional-Block Design of a Reactor
The channel-type reactor as a whole and its individual assemblies are now mainly
assembled at the construction site. The components of the reactor in the form of
parts or small assembly units are delivered from dffferent plants to the installa-
tion site, where they are assembled into a single structure. A large number of
spot assembly-welding operations with the necessary monitoring is carried out un-
der installation conditions in this case. It is understandable that the installa-
tion conditions are less adaptable for these types of operations than conditions
in specialized shops. Therefore, conversion of a large fraction of assembly and
welding operation~ from the installation site to plant conditions would make it
possible to incre~se the quality of reactor manufacture and to improve control and
increase its operating reliability. Moreover, this decision will provie a consid-
erable reduction of the periods of installation work in construction of AES and
consequently of the total construction period. All this yields a large saving.
~ ~
00 ' 00
3~ Q ~ ,
, Q ~
2-_ O , ',I 0 ,
I I . .
0 . ~ ,j~ 0
0 ~I ~ .
,
, , .
0 0 . .
- o------- o o.. o 0 0 0
- ; .
~oo a..~+ '
' o 4,r
Figure 11.1. Sectional-Block Reactor: 1-3--lower, side and upper blocks,
respectively
The reactor is manufactured in a specialized plant shop in the form of individual
blocks with all related assemblies and parts. The blocks are joined to each other
at the construction site of the AES during installation of equipment, forming a
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single structure [9). This reactor has been called a sectional-block reactor
(SBR). The core of an SBR is a rectangle in layout and is divided into individual
rectangular sections. The total number of reactor channels is determined by its
theYtnal output and the required number of channels for monitoring and control of
its operation. The number of channels in individual sections and consequently the
overall dimensions of sections and the number of sections are determined by the
capabilities of transporting the reactor blocks from the manufacturing plant to
the reactor installation site.
I
I
i
~ ~
~ .
,
Figure 11.2. Lower Block
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f
,
~
~
- I '
~ . ~
. ~ " '
- ~
~
Figure 11.3. Upper Block
An overall view of the reactor is presented in Figure 11.1. Each section is as-
sembled from lower, upper and side blocks. Making the core and consequently the
reactor in the shape of a rectangle with constant width permits one to increase
the reactor power to the required value by increasing its length by installing a
large number of standardized sections. (There is a limit in power in a cylindri-
- cal reactor, determi.ned by the maximum possible diameter of the upper plate). In
this case development of a new design of the main blocks and rearrangement of the
machine-building base are not required to convert to new, increased power. The
120
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central sections are designed Eor arrangement of the fuel and special channels and
the side reflector in them, while the two end sections of the reactor are used to
locate the end reflector in them.
The l.ower and upper blocks (Figures 11.2 and 11.3) are hollow sealed boxes in thC
form of parallelepipeds with internal longitudinal and transverse ribs to provide
stiffness to the structure, welded from sheet steel. The blocks are transported
in assembled form from the manu.facturing plant to the AES construction site on
special rail transport~rs (Figure 11.4). Transport by water, air or on truck-
trains is also possible.
24040 A-A .
q~ f000 1000
0
0
~ ~ ~h
- 1200 1200 ' o
A 1325Q 'Q'
~ 3fOf0
Figure 11.4. Transport of Reactor Block (third-order upper off-gauge load)
on Railroad Transporter
The sectional-block design of a uranium-graphite reactor permits an increase in
the quality of manufacture and at the same time an increase of operating reliabil-
ity and also considerably reduces the period of AES construction as a result of
conversion of a large part of operations in manufacture of the reactor from the
_ assembly site to specialized plant conditions. All this yields a large saving.
11.3. Steam Superheating in the Core
The turbines at most modern AES operate on saturated steam at pressure of 60-70
kgf/cm2 (approximately 6-7 MPa) in front of the check valve. These comparatively
low steam parameters are explained by the poor efficiency of existing zirconium
alloys used as the ma'r.erial of the fuel jackets at elevated temperatures and also
of the fuel channel pipes in channel-type RBMK reac~ors. At the same time the
efficiency of the AES is increased, capital expenditures for construction are re-
duced and the reliability an d maneuverability of the turbounits are increased
with stsam superheating in the core. Thermal discharges into the environment are
reduced and the flow rate of the cooling water through the turbine condensers is
reduced with an increase of efficiency.
If steam superheating to 450�C at presure of 65 kgf/cm2 (approximately 6.5 MPa) is
provided, then the thermal output of a reactor with superheating will be 11-12
percent less cnmpared to an energy unit with saturated steam of the same pressure
and same electrical output. This is a considerable advantage that leads to an
appreciable reduction of the nuclear Fuel consumption. The specific steam consump-
tion at the indicated superheating parameters are less by approximately a factor
of 1.4 than saturated steam consumption of the same pressure. The turbounit is
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improved, its reliability is increased and the operating conditions of the steam-
inlet and steam-outlet members of turbines are improved due to use of superheated
steam and a reduction of its consumption. A reduction of steam consumption also
results in a decrease of the number of main circulating pumps and separators in
the reactor unit.
All these advantages yield a considerable saving. However, an increase of coolant
parameters with introduction of nuclear superheating requires the use of materials
efficient at el~vated temperatures in the core.
_ ~1~
lkpeapemewl '
nvp ,
~ .
. A-A
~BBx4
_ 3
A II I A ' .
2 II ~
1
3
n�p~ (2)
Figure 11.5. Superheating Channel: 1--fuel assembly= 2--TVS hc~using; 3--
upper adapter; 4--zirconium pipe of channel= 5--lower adapter
Key:
1. Superheated steam 2. Steam
Water-graphite reactors of the Beloyarskaya AES in which steam is superheated in
- the core to 5Z0�C at pressure of 90 kgf/cm2 (approximately 9 MPa) have been oper-
ating for a long time in the Soviet Union (10]. Since the coolant temperature in
the superheated channels is considerably higher than that in the evaporation chan-
nels, the superheated fuel element jackets ahould no~v be made of steel. A zircon-
ium alloy should b~ used in the structure of superheated channels, as well as in
the evaporation channels, to manufacture the channel pipe. The fuel jackets must
first be made of steel since their temperature exceeds 600�C. This naturally re-
duces somewhat the neutron balance in the superheated part of the core.
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However, the advantage is comparatively low and the total saving achieved from
steam superheating remains significant when steel is used only for the fuel jackets.
The superheated channel (Figure 11.5) is similar in design to the evaporator chan-
nel and is a welded tubular structure 18-20 meters long. The channel is made of
zirconium pipe with outer diameter of 88 mm and wall thickness of 4 mm within the
core. The zirconium pipe is joined at the top and bottom to pipes of corrosion- ~
resistant steel by means of steel-zirconium adapters.
t~�C
600 , 5 I
, ' I
- 500 ~
~ ~ .I
_ 400 ~ 2 � I
~ g I
300 ~a 3
0 q02 QO~f p~ps .
RKO1C~ M�4�~C~KKQA (1)
Figure 11.6. Dependence of Steam Temperature and of Superheated Channel
Components on Thermal Resistance cf Fuel Assembly Housing:
1--steam temperature at aesembly output; 2--steam tempera-
ture at output from slit between assembly housing and chan-
nel pipe; 3--maximum temperature of inner surface of channel
pipe in heightt 4--maximum temperature of middle cross-sec-
tion of channel pipe in heiqhtl 5--maximum temperature of
outer fuel jacket in height; a--with housing in form of
72 X 1 pipe; b--with housing in form of two coaxial pipes
74 X 1 and 69 X 0.4 in diameter
Key :
1. m�hr��C/kcal
The characteristic feature of a superheated assembly design is the presence of a
housing. There is a slit 1-2 mm wide between the outer surface of the housing and
the inner surface of the channel pipe. As a result part of the saturated steam en-
tering the channel from below ia passed through the slit to coal the zirconium pipe
of the channel. The temperature dependence of the steam and superheated channel
components on the thermal resistance of the housing for thermal output channels of
2,400 kW with saturated steam temperature of 290�C at the inlet and superheated
steam temperature of 470�C at the channel output are shown in Figure 11.6. The
results were obtained at steam flow rate through the slit, comprising 25 percent
of the total flow rate through the channel equal to 14.5 t/hr.
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11.4. The Coolant Loop and Equipment Configuration
The sectional principle of reactor core configuration is feasible and can also be
u9pd in the design of the entire reactor unit. With this principle the coolant
circuit is separated into several lo~ops identical in output and composition ant~
equipment configuration. There are evaporation and su~erheated lonps in a reactor
with superheated steam in the core. The evaporation loop is a circuit of multiple
forced circulation, while the superheated loop is an open superheated steam cir-
cuit. The equipment of each loop is arranged in in~dividual boxes insulated from
adjacent boxes by protective walls, which permits repair operations in them with-
out complete shutdown of the reactor.
.
S , 6
~
4
B
3 9.
10
0
2 - f1 .
_ 12 .
~ ~ ~ 13
~ . .
� , .
. .
20 19 99 17 ' 16 15 14 . .
Figure 11.7. Configuration of RBMKP-2400 Reactor Unit (longitudinal section):
1--saturated steam collectort 2--superheated steam collector;
3--superheated steam pipelinest 4--saturated steam pipelines;
5--loading-unlaading machine; 6--steam separator; 7--assembled
group collectorsf 8--separated water pipelines= 9--steam-water
- mixture pipelines; 10--intake collector; 11--upper block; 12--
distributing group collectorsj 13--main circulating ptunp; 14--
delivery collectort 15--feed water collectorj 16--graphite
stackinqt 17--lower repair machine; 18--lower block~ 19--side
blocksj 20--bubbling basin
The multiloop principle of designing a reactor unit provides great advar:tages.
The entire unit seemingly consists of several relatively independent units. Be-
cause of this it has high flexibility. The individual loo~s or several loops can
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operate at lower power compared to the remaining loops or can be switched off
completely. This capability permits repair operations on the loop with the reac-
tor operatinq at reduced power both during planned repairs and during repairs
- CauSed by disruptions in operation of a given loop. The entire reactor must be
shut down in reactor units having no divisLon into individual loops located in
protective boxes to carry out any repair work. Therefor~, the repairability of a
sectional unit is higher than that of a nonsectional unit.
. ' ` '
, ~ ~ 3
~ . .
~ y
~ ~ - _
~ '
a ~~u
6 5
Figure 11.8. Configuration of RBMI~'-2400 Reactor Unit (transverse section):
1--steam separatorf 2--evaporation sectionsj 3--main circulat-
inq pump; 4--superheated loop boxt 5--superheated sectionst
6--evaporation loop box
Division of the unit into individual laops sharply increases its safety. The out-
put of each loop is relatively low. The diameters of all the pipelines and the
- dimensions of equipmenti are also considerably less than that without separation
into loops. Therefore, the consequences of an emergency will be considerably lower
if the seal of the circuit is broken due to ruptures of pipelines or equipment
failures. Thus, the measures required to ensure the safety of the environment and
population and also of plant personnel are easier to fulfill and require lower
expenditures.
The sectional-block design of a reactor and the multiloop version of the coolant
circuit permit easy variation of output of the power unit by selecting the re-
quired nimibex of standardized reactor sections and loops of the coolant circuit.
Additional research work is not required and new equipment does not have to be
produced in this case, which yields a large saving.
It is known that the unit output of power units will increase constantly. This
trend will also be maintained in the future. The problem of increasing the unit
outputs is solved comparatively simply with a sectional-block design. 2'he safety
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of the power units is not reduced in this case since it is determined by the char-
acteristics of the individual loop rather than by those of the reactor and the cir-
cuit as a whole.
' O~
, ~ .
~ ~
m '
~8 ~
gg9 , '
Figure 11.9. RBMKP-2400 Reactor Unit
The configuration of the RBMKP-2400 reactor unit, consisting of a sectional-block
reactor with nuclear superheating of steam and multiloop coolant circuit, is pre-
sented in Figure 11.7, 11.8 and 11.9.
Channel-type power reactors with boiling-water coolant have achieved considerable
development in the Soviet Union. There is considerable positive experience of
their reliable operation at existing AES, upong which the first--the Leningrad AES
imeni V. I. Lenin--has been operating successfully since 1973 (Figure 11.10).
There are also great prospects for a further significant improvement of AES with
these types of reactors. A reactor with ~mproved engineering and economic indi-
cator�s--the RMBI~-2400--has now been developed. A graph of an atomic power plant
with RBMKP-2400 type reactor is shown in Figure 11.11.
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a~, _ . , r , ,
, ~ , *t i ,
' R ; ~ , . . f . . , ' . . . .;r ,'i
. ,
. . ' ' . ~ ~ . . ~ , ~
~ . . ~ . . . ' ' . ~
..~I._ ~ . t,: ~
~i~~~ tw ~ ~ ' , Gt+
9 s~~~ (
! A~~~~? / C 1u.
~ V ,
- h:S,X .',P�+RtpL . .
f lrw~r...
i"
j,:~r..~'w
~
�~y'. 4 ~:i.
k �
Figure 11.10. Leningradskaya AES imeni V. I. Lenin (first unit)
, . . .
. ,
~
;~~~-=_C ,
~ ~ ~ t
.rwi;
~ _ + ~
~ +
: ~
~t.
r
~
_ I ~Ci~l'r,tr~ !"~y \y
..~~~1`~:.~; ~ ' .
~ ~ fi . ~
~.~ti 'a''r ~'ij~%i.
1�~1. 11t ~n� r ~ .~~ia
:�{7~f
~ ~ ' ~L ~I'.~~}~ ~ T:.. . �i~(~ej'~ _ b
'S~at - ~,Gr,..;~i ~ C~ .M: 1%i.:. 4= . M ` .)d_'
,~"'~7fF ~ s p�c~. .
ti . .nti ~lK!~'rt ~ ~ ~'.~r`.~.~:.~:s' ,
Figure 11.11. Overall View of Planned AES with RBMI~-2400 Reactor
BIBLIOGRAPHY
1. Petros'yants, A. M., "~ovremennyye problemy atomnoy nauki i tekhniki v SSSR"
(Modern Problems of Atomic Science and Engineering in the USSR~, Moscow, � ,
Atomizdat, 1976.
2. "Atom sluzhit sotsializmu" [The Atom Serv~s Socialism], Moscow, Atomizdat,
1977.
127
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3. Aleksandrov, A. P., Yu. M. ~ulkin, I. D. Dmitriyev et al, "Physical and Energy '
Startup of the First Unit of the Leningrad AES imeni V. I. Lenin," ATOMNAYA
ENERGIYA, Vol 37, No 2, 1974.
4.~ Yemel'yanov, I. Ya., P. A. Gavrilov and B. N. Seliverstov, "Upravleniye i
bezopasnost' yadernykh reaktorov" [Control'and Safety of Nuclear Reactors],
Moscow, Atomizdat, 1975.
5. Dollezhal', N. A. and I. Ya. Yemel'yanov, "The Experience of Developing Large
Power Reactors in the iJSSR," ATONIIdAYA ENERGIYA, Vol 40, No 2, 1976.
6. Aleksandrov, A. P. and N. A. Dollezhal', "Development of Channel-Type Uranium-
Grahpite Reactors in the USSR," ATONBJAYA ENERGIYA, Vol 43, No 5, 1977.
7. Aden, V. G., Yu. K. Bibilashvilli, A. S. Zaymovskiy et al, "The Fuel Element
of the RBMK-1000 Reactor," ATONII~JAYA ENERGIYA, Vol 43, No 4, 1977.
8. Tsykanov, V. A. and Ye. F. Davydov, "Radiatsionnaya stoykost' teplovydelya-
yushchikh elementov yadernykh reaktorov" (The Radiation Stability of the Fuel
Elements of Nuclear Reactors], Moscow, Atomizdat, 1977.
9. Dollezhal', N. A., I. Ya. Yemel'yanov, Yu. M. Bulkin et al, "Channel-Type �
Sectional-Block Reactor with Nuclear Superheating of Steam with Electrical
Output of ~ao Million kW," in "Opyt ekspluatatsii AES i puti dal'neyshego ~
razvitiya atomnoy enerqetiki" LZ'he Experience of Operating AES and Methods of
Further Improvement of Atomic Power Ehgineering], Vol 2, Obninsk, FEI, 1974. ,
10. Dollezhal', N. A., V. M. Malyshev, S. V. Shirokov et al, "Soane Results of ,
Operating the Beloyarskaya AES imeni I. V. Kurchatov," ATONAJAYA ENERGIYA,
Vol 36, No 6, 1974.
~
11. Yemel'yanov, I. Ya., A. D. Zhurnov, V. I. Pushkarev and A. P. Sirotkin, ~
"Increasing the Efficiency of Uranium Utilization in the RBMK-1000," ATONIDiAYA
ENERGIYA, Vol 46, No 3, 1971.
COPYRIGHT: Atomizdat, 1980 ,
6521
- CSO: 8144/0120
_ ;
- END -
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