SOVIET ATOMIC ENERGY VOL. 57, NO. 2

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• Publication Date: August 1, 1984
• Content Type: REPORT

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Russian Original Vol. 57, No. 2, August, 1984
February, 1985
SATEAZ 57(2) 507.576 (1984)
SOVIET
ATOMIC
ENERGY
ATOMHAH 3HEPi'NR
(ATOMNAYA ENERGIYA)
TRANSLATED FROM RUSSIAN
CONSULTANTS BUREAU, NEW YORK
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SOVIET
ATOMIC
ENERGY
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
February, 1985
Volume 57, Number 2 August, 1984
CONTENTS
Engl./Russ.
ARTICLES
An Experimental Study of Emergency Cooling Conditions in RBMK Reactors
on Load Disconnection - V. N. Smolin, V. I. Esikov, V. P. Shishov,
S. P. Kuznetsov, and V. S. Grigor'ev. . . . . . . . . . . . . .
507
83
Safety under Servicing Cooling Conditions for RBMK Reactors
- V. N. Smolin, V. I. Esikov, V. P. Shishov, V. P. Vasilevskii,
and V. S. Grigor'ev . . . . . . . . . . . . . . ? , . ,
512
87
Gas Phase in Experimental Fuel Elements with Compact Uranium Dioxide,
Irradiated in the Sm-2 Reactor - A. P. Kirillovich,
V. Sh. Sulaberidze, Yu. I. Pimonov, V. N. Shulimov,
Yu. G. Lavrinovich, A. S. Biryukov, and V. N. Kupriyanov. . . . . .
517
91
Influence of the Texture of Prismatic Planes on the Anisotropy
of Deformation of Irradiated Zirconium Alloys - Yu. N. Knizhnikov
and V. V. Kolomytkin . . . . . . . . . . . . . . . . . . . . . . . .
523
95
Ion- and Electron-Stimulated Low-Temperature Desorption of Gases
Dissolved in Metals - N. P. Katrich and V. N. Kanishchev.. . . . .
527
99
Absorption Parameters of Deuterium Ions in Molybdenum - A. A. Pisarev
and V. N. Tsyplakov . . . . . . . . . . . . . .
534
104
Dehydration of the Steam-Generating Channel Due to Cooling Loop
Depressurization - E. I. Liverant, A. P. Proshutinskii,
and E. M. Staviskii . . . . . . . . . ? . . . . .
540
108
Studying the Electrophysical Parameters of Piezoceramics of Various
Types in an IVV- 2M Reactor - Yu. P. Meleshko, S. V. Babaev,
S. G. Karpechko, V. I. Nalivaev, Yu. A. Safin, and V. M. Smirnov.
544
ill
Use of Monoisooctylmethylphosphonic Acid and.Its Trivalent Iron Salt
in Determining Radionuclides in Effluents - N. E. Tsvetaeva,
V. M. Filin, L. A. Ivanova, V. N. Revnov, E. P. Rodionov,
L. Ya. Rudaya, I. A. Suslin, and K. Yu. Shapiro . . . . . . . . . .
548
114
A Method of Calculating Membrane-Element Cascades for Separating
Multicomponent Mixtures - E. B. Gruzdev, N. I. Laguntsov,
B. I. Nikolaev, A. P. Todosiev, and G. A. Sulaberidze . . . . . . .
(55 7'
117
LETTERS
Density of Melts of the Ternary Mutual System K, Kr//F, Cl
- S. E. Darienko, N. N. Kurbatov, S. P. Raspopin,
and Yu. F. Chervinskii. . . ... . . . . . . . . . . . . . . . . . . .
558
122
Heat Exchange during the Flow of a Melt of LiF-NaF-KF Fluoride Salts
in a Circular Tube - V. V. Ignat'ev, S. V. Keronovskii, -
A. I. Surenkov, 0. P. Shcherbanyuk, S. P. Manchkha,
and Yu. B. Smirnov. . . . . . . . ? . . . . . . .
Effect of Reactor Irradiation on the Microstructure of Pyrocarbon
Coatings - I. S. Alekseeva, A. A. Babad-Zakhryapin.,
Yu. G. Degal'tsev, L. A. Elesin, Yu. M. Utkin,-and Yu. N. Yurovskikh.
562
124
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CONTENTS
(continued)
Engl./Russ.
High-Speed Bulk Analysis of Uranium in Geological Samples by the Delayed
Neutron Method - V. R. Burmistrov . . . . . . . . . . . . . . . .
565
126
Isotopic Composition of Cesium Built Up in VVER Nuclear Fuel
-V. Ya. Gabeskiriya, V. I. Borisenkov, V. V. Kalygin,
V. M. Prokop'ev, V. S. Prokopenko, M. N. Maslennikova,
and A. P. Chetverikov . . . . . . . . . . . . . . . . . . . .
567
127
Saturation of the Surface Layer of Carbon Devitrified Glass, Boron
Nitride, and Quartz Glass with Ion-Implanted Deuterium
- V. V. Bandurko, V. V. Kulik, A. A. Pisarev, and V. N. Tsyplakov . .
569
128
Yields of 42K and 43K Upon Irradiation of Calcium by Protons and Deuterons
- P. P. Dmitriev and M. V. Panarin . . . . . . . . . . . . . . . .
571
130
Implantation of Low-Energy Hydrogen Ions in Lithium - 0. G. Voronkov,
V. F. Zubarev, and L. M. Frantseva. . . . . . . . . . . . . . . . .
573
131
The Russian press date (podpisano k pechati) of this issue was 7/26/1984.
Publication therefore. did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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ARTICLES
AN EXPERIMENTAL STUDY OF EMERGENCY COOLING CONDITIONS IN RBMK
REACTORS ON LOAD DISCONNECTION
V. N. Smolin, V. I. Esikov, V. P. Shishov, UDC 621.039.524.4
S. P. Kuznetsov, and V. S. Grigor'ev
When the internal equipment in a nuclear power station is deprived of current, the main
circulation pumps (MCP) stop along with the feed pumps (FP), while the emergency shutdown (ES)
equipment operates and the automatic shutoff valves ahead.of the turbines are closed, which
causes the pressure to.increase and the safety valves to open. Then the pressure in the loop
begins to fall, and the safety valves should close. After about 3.min, the emergency feed
pumps (EFP) are switched on. It has been found with a simulation system and checked on the
reactor that stable conditions are then set up in the loop by natural circulation (NC), and
cooling the core does not cause any complications.
However, if one or more of the safety valves do not close properly, the pressure in the
loop will continue to fall, and the rate of fall and the final value are dependent on the de-
gree to which the safety valves fail to close. The rate of fall and the extent of the pres-
sure fall influence the circulation parameters and the fuel-element cooling conditions, and
the extent of these effects is examined here. Because it is very difficult to make analyti-
cal calculations for. such conditions, one needs experiments on models. A test model for the
RBMK has been built at the Power Engineering Research and Development Institute (V. N. Smolin
et al., Voprosy Atomnaya Nauki i Tekhniki, Ser. Reaktorostroenie, Issue 1(8), 3 (19.74)). This
simulates closely the forced multiple circulation loop (FMCL) in a reactor, including the
positions at the start and end of the hydraulic system and the main equipment. This facility
provides hydraulic resistances and coolant circulation times in the incoming and outgoing
pipes corresponding to those in the actual reactor. The upper and lower positions in the dis-
tributing group collectors (DGC) are simulated by means of appropriate loop sections. The
volume and surface of the separator per fuel element are similar in the model to those in the
reactor.
Figure 1 gives the cross sections of the loops in the model and reactor in relation to
the relative loop length per fuel element (the experimental assembly contained seven heated
rods). The model retained virtually the same relative loop volume per fuel element as in the
reactor (0.045 and 0.044 m3, respectively). The height distributions of this volume were al-
so identical. The test methods involved reproducing the operations and circulation condi-
tions in the model with the maximum possible similarity to the actual processes.
The initial state was that of stationary NC at a pressure of about 7.0 MPa, since the
previous changes in loop parameters do not influence the subsequent processes. We also re-
corded the coolant flow rate through the experimental channel (EC), the feedwater flow rate,
the pressure and level in the separator, the EC power, the coolant temperatures at the inlet
and outlet to the heat-production zone, the temperatures of the fuel-element simulators in the
.EC, and the pressure differences across components of the circulation loop.
The model was used to examine emergency cooling conditions not only by means of EC when
the feedwater supply to the separator is restored but also by means of forced circulation in
a direct-flow scheme with the feedwater supplied directly to the DGC. To define criteria and
parameters governing the occurrence of hazardous conditions, the restored flow rate for the
feedwater was different from that provided by the EFP in certain experimenta. The end of the
state with NC was taken as the time when the working conditions stabilized with a given final
pressure in the separator and normal cooling in the heat-production zone or the point of fuel-
element overheating. The purpose of the NC experiments was to determine the final pressure
and the other circulation parameters for various rates of pressure reduction leading to fuel-
element overheating. The end of the state with forced circulation was taken as the instant
when the heat-production zone was cooled below the saturation temperature with the given rate
Translated from Atomnaya Energiya, Vol. 57, No. 2, pp. 83-87, August, 1984. Original
article submitted February 10, 1984.
0038-531X/84/5702-0507$08.50 ? 1985 Plenum Publishing Corporation 507
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IIYMTMTH
Fig. 1. Comparison of the cross sections
of the model (1) and reactor (2) per fuel
element by reference to the reduced height.
MW F
P
s
50
2,5
0
N kw 450
300
750
Gf W':
kg /h '00
IG, kg/h 71700
12171117
1000
L0700[
0
4 MPa2
00
200. 0 2 4 ; miff
Fig. 2. Change in parameters in emergency cooling of the heat-
production zone: 1) lower position of the DGC; 2) upper position
of the DCG; ps) pressure in.separator; N) channel power; Gfw)
feedwater flowrate; G) circulation flow rate; L) level in separa-
tor;,Ap) pressure difference in outlet section; tfe) fuel-element
temperature; and the hatched region represents flow-rate pulsa-
tions.
of pressure fall and given feedwater flow rate or the instant of fuel-element overheating.
The purpose of the experiments in the second method was to determine the minimum permissible
pressure that does not lead to disruption of the normal cooling in the heat-production zone
as affected by the reduction rate as coolant circulates from the EFP.
The initial state was always the same in all the experiments: the stationary state of NC
with a pressure of 6.7 MPa in the separator, power 0.3 of nominal, average coolant tempera-
ture at the inlet to the heat-production zone 270?C, and coolant flow rate 0.35 of the nom-
inal value. The average pressure reduction rate was 0.2-1.5 MPa/min, while the final value
of the pressure was 0.68-4.8 MPa. The power was reduced to 0.1 of the nominal value. At the
instant when the power was reduced and the pressure began to fall, the feedwater supply to
the loop was halted, and then after 3 min it was restored with a flow rate of 0.1 of the nom-
inal value.
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2,0
2000
4 kg
0
300
N,kgh
0
Gfw,kg h0
200
0
% 100
50
p. MPa
? ^ ?
300
0,2 0,4 0,0 0,8 Dp -MPa/min
MPad
0,04
fe,0C
200
300
0
in' C 250
Fig. 3 Fig. 4
Fig. 3. Dependence of final pressure in the separator corresponding to normal cool-
ing in the heat-production zone ( 0 and 0 ) or fuel-element overheating ( ? and
^ ) on pressure reduction rate; 0 , ? ) upper position of DGC; 0 , ^ ) lower position
of DGC; dashed line) limiting pressures,for lower position of DGC (1) and upper posi-
tion (2).
Fig. 4. Parameter variation in emergency cooling of the heat-production zone by
forced circulation from the ECP with feedwater supplied to the DGC: tin) tempera-
ture at the inlet to the zone, Appz) pressure difference in the heat-production
zone; for other symbols, see Fig. 2.
.Figure 2 shows the typical behavior of the parameters during the transient states in NC
accompanied by fuel-element overheating. When the feedwater is supplied with a flow rate
close to the output of the ECP, fuel-element overheating sets in at the same separator pres-
sure, no matter what the rate of pressure reduction. However, the final pressure is dependent
on the FMCL scheme: It was 4.5 MPa with the upper position of the DGC or 3.8 MPa with the
lower. one (Fig. 3).
The experiments showed that if the loop is supplied with feedwater in an amount greater
than that from the ECP, the final pressure is reduced. If the feedwater is supplied at a
flow rate sufficient to maintain the pressure difference in the outlet part no less than 12 m
of water column (this is close to the nominal feedwater flow rate in the model as reckoned
for 100% reactor power), then a stable NC condition is set up in the loop, which does not
lead to deviation from the normal cooling in the heat-production zone as the pressure falls
in the range examined.
As the hydraulic system in the model simulates the FMCL closely, the actual changes in
NC parameters on emergency cooling should be analogous and the experimental results should be
0
300
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7,0
Psi MPa
5,0
3, 0
2000
G,.kg /h
1000
0
300
N, kw
0
400
Gfw. kg/h
200
0
100
50
0.08
APPz, MPa
0,04
P. MPa
50
0 0
o 00 o
200
i
0L 0,25 0,5 0,75 tO,,MPa/min 150 0 8 18 .24 -r, min
Fig. 5 Fig. 6
Fig. 5. Dependence of the minimum permissible pressure in the separator on emer-
gency cooling by the supply of feedwater from the ECP to the DGC as affected by pres-
sure reduction rate: 0 ) state with normal cooling in the heat-production zone; 0 )
state with fuel-element overheating.
Fig. 6. Parameter variation in emergency cooling in.the heat-production zone by
forced circulation from the ECP with feedwater supply to the DGC. See Fig. 2 for
symbols.
transferable directly to the reactor. To confirm this conclusion, we determined the separa-
tor filling times in the model and on the reactor, since this time. is a decisive factor in
these processes. We took a condition in which the pressure fell at 0,5 MPa/min (Fig. 2).
With the coolant flow rate corresponding to the initial period under these conditions, the
saturated water from the separator filled all the outlet and inlet sections, and the level
in the separator began to rise at the end of the third minute. We determined the steam gen-
eration rate due to spontaneous boiling and the rate at which the steam was removed from the
loop, and these data were used to estimate the rate at which the separator filled with water.,
The calculated separator filling time was 95 sec for the model at a pressure of 5.2 MPa, while
the values for the reactor were 95 sec and 5.1 MPa, so there is good agreement between these
values and also with the experimental data.
Therefore, if the pressure in the RBMK circulation loop falls below the saturation pres-
sure because the safety valves fail to close and the loop is receiving an inadequate feedwa-
ter supply, then the coolant boils in the outlet part, which causes the separator to overfill,,
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MPa/min
0,2 Gfw / ufwm
Fig. 7. Dependence of the permissible pressure reduction rate
on the feedwater flow rate supplied to the DGC: 10 ) condition
with normal cooling in the heat-production zone;'e ) condition
with fuel-element overheating.
while coolant is lost from the loop, and consequently there is a reduction in the NC flow rate
through the core, while the fuel elements may overheat. To prevent this, it is necessary to
supply the loop at a certain instant with water in an amount sufficient to maintain a heat
of not less than 12 m of water column in the outlet pipe.
We examined the core cooling by forced circulation in the same sequence. At 3 min after
the simulation of the emergency, which corresponded to power reduction and pressure fall at a
given rate, feedwater was supplied directly to the DGC in an amount simulating the activation
of the ECP. At the same time, the cutoff valves isolated the outlet part of the loop (this
simulated the operation of the check valves after the MCP).
Figure 4 shows the parameter variation under one such set of conditions with the pressure
falling at 0.3 MPa/min. Figure 4 shows that the parameter change is as in NC states up to
the end of the third minute, and then the fall in pressure is accompanied by coolant loss in
the form of a steam-water mixture after the feedwater is supplied to the DGC and the outlet
part has been cut off, which is indicated by the ongoing overfilling of the separator and the
monotonic fall in the pressure in the heat-production zone. Under these conditions, the nor-?
mal cooling of the heat-production zone is disrupted at 2.3 MPa. It was characteristic that
the feedwater front did not reach the heat-production zone, as was evident from the coolant
temperature at the inlet to it, which was equal to the saturation temperature at the inlet
pressure.
The experiments showed that for a given feedwater flow rate, the minimum permissible
pressure increases with the pressure reduction rate (Fig. 5). Under all these conditions, it
is characteristic that the coolant temperature at the inlet to the core is approximately equal
to the saturation temperature. However, while the coolant temperature at the inlet to the
heat-production zone remained below the saturation temperature until the minimum pressure was
reached, which corresponds to that temperature, there was no fuel-element overheating. That
condition can be provided by increasing the feedwater flow rate.
Figure 6 shows one of these cooling conditions with the loop pressure falling. It is
evident that the fuel-element cooling by the feedwater having a temperature below the satura-
tion value occurred at minute 18 and a pressure of 3.5 MPa with the pressure falling at a
rate of about 0.2 MPa/min, that pressure exceeding the permissible value by 1.0 MPa. The on-
going pressure reduction in the loop and the stabilization at the level of 1.0 MPa did not
cause fuel-element overheating. It was also found that the safe rate of pressure reduction
increased with the feedwater flow rate to the DGC (Fig. 7).
As the hydraulic system and pipeline sections and lengths in the FMCL in the model cor-
responded completely to the FMCL in the reactor, we conclude that the circulation parameters
should also be analogous, and therefore the experimental results on emergency cooling with
feedwater supply to the DGC can be transferred directly to the reactor.
Pressure reduction in the FMCL of the reactor at rates of 0.2-1.5 MPa/min with a 3-min
interruption in the feedwater, which is subsequently supplied to the separator with a flow-
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rate of 10% of the nominal value, leads to the coolant boiling in the loop, which raises the
water level in the separator, with water loss from the loop, and with a fall in hydrostatic
head in the outlet part and disruption of the normal cooling in the heat-production zone when
the pressure falls below the permissible value (3.8 MPa with the lower position for the DGC,
or 4.5 MPa in the upper one).
If the feedwater is supplied to the separator at a flow rate sufficient to maintain the
hydrostatic head in the outlet part at not less than 12 m of water column (this flow rate in
the model corresponded to the nominal feedwater flow rate at 100% reactor power), then a sta-
ble NC condition is set up in the loop, which does not lead to disruption of the normal cool-
ing in the heat-production zone even at pressures below the permissible value.
Emergency cooling by supplying feedwater to the DGC provides for normal.cooling of the
heat-production zone with larger falls in pressure than does cooling by the use of NC. If the
feedwater is supplied in an amount providing for the coolant to enter the heat-production zone
at a temperature below the saturation point, then normal fuel-element cooling is provided when
the pressure falls below the minimum permissible value.
Activation of the ECP and feedwater supply to the DGC with a flow rate of 10% of the
feed-pump output provides for normal cooling of the heat-production zone at pressure fall
rates up to 0.2 MPa/min. To provide safety with emergency cooling at pressure reduction rates
greater than 0.2 MPa/min, it is necessary to supply the DGC with a larger amount of.feedwater
from other sources in order that the heat-production zone should receive water at a tempera-
ture below the saturation point before the minimum pressure is.reached.
SAFETY UNDER SERVICING COOLING CONDITIONS FOR RBMK REACTORS
V.
N.
Smolin, V. I. Esikov,
V.
P.
Shishov, V. P. Vasilevskii,
and
V.
S. Grigor'ev
The RBMK-1000 graphite-channel boiling-water reactor has 1693 fuel channels (FC) arranged
in vertical holes in the graphite stack. Each channel is a body of tubular construction with-
in which are located two fuel-element assemblies each containing 18 fuel elements which are
flushed on the outside surface by the coolant [1].
The residual power production in the core when the reactor has been shut down is fairly
substantial. For example, after a day it is 0.4% of the nominal power Nnom, i.e., 12.8 MW.
After 30 days, this falls to 0.12% of Nnom and then remains virtually constant for a long time.
This makes clear why it is not permissible to drain the core even after shutdown. Therefore,
in conducting servicing on the forced multiple circulation loop (FMCL), it is necessary to
organize core cooling.
One of the basic specifications for such a cooling system is that the cooling should be
reliable and that safe access should be provided to the FMCL for examination or repair. This
is attained by installing shutoff valves at various parts of the circuit to provide for drain-
ing, and also for organizing various core cooling modes.
During the design of the RBMK, three conditions of service cooling were provided in order
to facilitate servicing (the corresponding schemes are shown in Fig. 1):
1) natural circulation with nominal water levels in the separators and the FMCL valves
open;
2) interrupted natural circulation with the separators drained and the FMCL valves open;
and
3) bubble mode with nominal water levels in the separators but with the pressure-regulat-
ing valves (PRV) at the inlet to the FC closed.
Translated from Atomnaya nergiya, Vol. 57, No. 2, pp. 87-91, August, 1984. Original
article submitted February 10, 1984.
0038-531X/84/5702-0512$08.50
O 1985 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/02/22 : CIA-RDP10-02196R000300050002-5
Fig. 1 Fig. 2
Fig. 1. Schemes for cooling the core in FMCL servicing: a) natural circulation; b)
interrupted natural circulation; c) bubble mode; 1) valves open; 2) valves closed;
HFMCL) water level in FMCL.
Fig. 2. Volumes of hydraulic systems in relation to length for an experimental chan-
nel (1) and a fuel channel (2) as referred to one fuel element.
Natural circulation in the FMCL is provided by cooling the water in the outlet section
of the loop, for which one uses theordinary flushing and cooling system. The coolant heated
in the core is transferred by the cooling pumps from the water tapping points in the separa-
tors to the flushing cooler., where the temperature is reduced by heat transfer to the water
in the intermediate circuit. The cooled water passes to the feed pipelines and then to mix-
ers at the inlet, which thereby cools the circuit. This condition is used for. ordinary reac-
tor cooling, in repairing the main circulation pumps (MCP), and also in servicing.the.pres-
surized and suction pipes, as well as for preliminary cooling of the reactor and-FMCL before
the start of servicing. -
In the state of interrupted natural circulation, the separators are drained and commun-
icate with the atmosphere. The core is supplied by spontaneous flow from a servicing tank
connected to the pressurized collectors in the FMCL by special pipes. The gate valves in the
MCP may be closed, i.e., it is possible to drain the FMCL as well. In this state, one can
service the separators, the pipelines, the suction collectors, and the MCP pipelines with
their valves. Here, to provide for safety, special rubber-metal plugs are inserted in the pipe-
lines from the collectors and the MCP.
In the third state, the PRV at the inlet to the FC are closed, and the level in the sep-
arators is nominal. Under these conditions, one can repair the equipment and the FMCL pipes
on the section from the inlet gate valves on the MCP to the PRV. This mode of cooling is
widely used in general replacement of failed transducers in the flowmeters and PRV. In the
latter case, a special freezing system provides ice plugs in the water pipelines.
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I tfe- *C -- -----
100
sn
itin- 'C F__
3
S00_ 10 20 30 40 % min
Fig. 3. Time variations in the temperatures of fuel-element
simulator (1), of the coolant at the.channel inlet (2), and
at the outlet (3) with a water level in the FMCL above the
horizontal part of the SWP.
Before implementation on the RBMK, the modes of core cooling were examined with a model
at the Power Engineering Research and Development Institute. A detailed description of the
model has been given in [2], which includes simulation of the FMCL. The lengths and cross
sections of the water pipes have been made such that the hydraulic resistances, circulation
times, and volumes of the various parts per fuel element corresponded to the FMCL in the re-
actor (Fig. 2), and the same applied to the heat-production zone, the exit part of the chan-
nels, and the steam-water pipes (SWP). The mutual dispositions of the collector, heat-produc-
tion zone, channel outlet, and separator in the model were also identical with those in the
FMCL in the reactor. The heat-production zone in the experimental channel was a bundle of
seven rods simulating fuel elements of length 7 m and outside diameter 13.5 mm, which were
arranged in a triangular lattice with a pitch of 16 mm. The heat was produced by passing di-
rect current through the rods. The simulation principles led us to believe that the hydrody-
namic and thermophysical processes in the hydraulic system of the model would be identical
with those in the fuel channels.
Tests with the model showed that one got the usual natural circulation in the FMCL with
the separators filled with water to the nominal level and the PRV open, and this provided
reliable cooling at any pressure.
When the water level falls below the ends of the SWP tubes with the PRV open at atmospher-
ic pressure in the separator, one gets interrupted natural circulation, in which there is no
bubbling of the steam formed in the heat-production zone through the layer of water in the
SWP. In this mode, the water is periodically ejected by the steam from the upper part of the.
channel and the SWP into the separator and is drained into the descending part. The differ-
ence in hydrostatic pressures also causes periodic entry of water from the descending part
into the heat-production zone. This also provides reliable channel cooling if the steam.
ejected from the separator is compensated by supplying the loop with water and the water lev-
el in the FMCL is above the horizontal part of the SWP (Fig. 3).
When the water level in the FI4CL falls below the horizontal part of the.SWP (with the.
PRV open), the picture is very different. In that case, the steam displaces the water from
the volumes above the heat-production zone into the upper part of the channel and then into
the SWP. Here again, water from the descending section flushes the heat-production zone.
However, this process continues until the horiz...ontal part of the SWP and some of the vertical
part have filled with water to equalize the hydrostatic pressures in the descending pipes and
the channel. Then the filling of the heat-production zone with water ceases, and the water
boils, so the temperatures of the simulation rods increase considerably (Fig. 4). This shows
that it is impermissible for the water level in the FMCL to fall below the horizontal part
of the SWP even with the valve at the inlet to the channel open.
We examined the cooling in. the FC with the valve at the inlet closed at atmospheric pres-
sure in the separator or with an excess pressure in it (up to 0.6 MPa), while the power level.
in the experimental channel corresponded to 0.8-1.7% of the power in a maximally loaded FC.
In the first case, the separator communicated with the atmosphere, while in the second the
separator was supplied with compressed air.
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 Declassified and Approved For Release 2013/02/22 : CIA-RDP10-02196R000300050002-5
t out,
500 20 40 6/1 BO [. min
Fig. 4
1 12
8
4
I bsep'
i - T--r
MW
I I
"''I I I
i I
tSep. *OC -c --~~
tfe- C
120
100 0 9 7 3 4 5" 7, min
Fig. 5
Fig. 4. Time variations in the temperatures of fuel-element simulator (1), of the
coolant at the channel inlet (2), and at the outlet (3) with a water level in the
FMCL below the horizontal part of the SWP.
Fig. 5. Variations over time in the parameters of the hydraulic system in the model
at atmospheric pressure in the separator with acceptable channel power. Hereand
in Fig. 6: Hpz, Hucs, Hswp, and Hoa are the pressure differences in the heat-produc-
tion zone, upper channel section, steam-water pipe, and in the system overall, while
hsep is the water level in the separator, and tsep and tfe are the temperatures in
the separator and in a fuel element.
The changes in hydrostatic head in various parts of the channel indicate the state of
the coolant in the loop and the processes occurring there under conditions of reliable cool-
ing with atmospheric pressure in the separator (Fig. 5). At the instant corresponding to
point a, the entire hydraulic system in the channel is filled with water from the separator.
During time a-b, the water in the core is heated to the saturation temperature, as is. evident
from the stabilization of the saturation temperature at a pressure equal to the hydrostatic
head at the level where the thermocouple lies. Point b corresponds to the start of bulk
boiling in the core, as is evident from the reduction in hydrostatic pressure not only in the
core but also generally...- The retention of unchanged pressures in;the upper channel section
.UCS above the core and in the SWP in period b-c indicates that the steam does not bubble
through the water in the hydraulic system above the core, but instead the growing steam bubble
displaces the water into the free volume of the separator (see the water level in the separa-
tor in Fig. 5). The period c-d corresponds to the steam displacing the water from the UCS,
while d-e corresponds to displacement of the water from the horizontal part of the SWP (the
hydrostatic pressure in the lifting part of the SWP remains unchanged). During period a-f,.
the steam displaces the water from the vertical part of the SWP, and at time g the steam
reaches the separator, at which point the coolant temperature at the inlet to the separator
becomes equal to the saturation temperature. When the steam enters the separator, it con-
denses, which produces avalanche filling of the entire hydraulic system with water from the
separator (point g in Fig. 5). This figure shows that during the period b-g the separator
gradually fills with water and then suddenly empties to flush the circuit. The process re
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 Declassified and Approved For Release 2013/02/22 : CIA-RDP10-02196R000300050002-5
HpZ, a b c d e
~ n
Hucs,
MW
Fig. 6. Time course of the parameters in the hydraulic system
in an experimental channel. at atmospheric pressure in the sep-
arator with channel power above the permissible level.
peats in subsequent cycles. A characteristic feature is that the pressures in parts of the
circuit in a given part of the cycle attain the same values as in the previous cycle. This
cyclic process persists for any length of time for a constant channel power and constant wa-
ter temperature in the separator.
.The experiments showed that the period of this cycle is dependent on the channel power,
decreasing as the.power increases. Also, if the water temperature in the separator is below
70?C, there are water-hammer effects during the filling of the channel with water and the con-
densation of the steam there.
When a certain channel power is exceeded, the course of the process alters (Fig. 6).,The
hydrostatic pressures in the channel as a whole and. in the individual parts in each succes-
sive cycle are reduced relative to the previous cycle, which indicates reduced water entry
into the hydraulic system. Finally, the heat-production zone becomes almost entirely free
from water, and then the steam begins to be superheated and the temperatures of the separator
rods rise to an impermissible level.
When there is excess pressure in the separator, the process is very different. When the
valve at the channel inlet is closed, the water in the heat-production zone is heated to the
saturation temperature, and then steam begins to form. The resulting steam bubble increases
in volume and displaces the water from the hydraulic system into the separator, thus filling.
part of the heat-production zone, the upper channel section, and part of the SWP. In all.
such experiments, the process was accompanied by steam superheating in the core and a sub-
stantial increase in the rod temperature at any pressure exceeding atmospheric and at power
levels corresponding to the FC power from 20 to 50 W. It seems that when there is excess
pressure in the separator, the bulk steam generation in the core is less, and it is insuffi-?
cient to displace the water from the entire hydraulic system. Only in that case could the
steam reach the separator, where it would condense and the system would be filled by water.
Therefore, this study of the FC cooling conditions with the inlet valve closed has shown
that reliable element cooling can be provided with atmospheric pressure in the separator by.
the hydraulic system being filled periodically with water from the separator. The maximum
permissible power level indicated by the studies on the model is 25 kW for the real FC.
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 Declassified and Approved For Release 2013/02/22 : CIA-RDP10-02196R000300050002-5
These experiments with the model on the cooling of the RBMK core have led us to formulate
technological specifications covering safety during reactor operation if the circulation ceas-
es and parts of the FMCL are drained, which have been inserted in "Engineering Rules for Oper-
ating Nuclear-Power Station Units Containing RBMK Reactors."
1. Transfer of the FC to bubble mode, i.e., closing the valves at the inlet, is allowed
not earlier than-72 h after shutting down the reactor in order to reduce the residual power
to a permissible level if the following conditions are met:
a) water levels in the separators above the ends of the upper series of SWP;
b) the water temperature in the separators should be not less than 80-90?C in order to
avoid hydraulic shocks in the SWP; and
c) the pressure in the separator is atmospheric.
2. It is forbidden to reduce the water levels in the separators below the ends of the
SWP if the FC have closed inlet valves.
3. In all cases where parts of the FMCL are drained, the water level should not fall
more than 1 m below the heads of the FC, and then it is necessary to provide a constant supply
of cold water to the core. Also, studies on servicing conditions involving water level reduc-
tion in the loop have enabled us to establish the limiting positions for the upper and lower
water levels in the servicing tank, which is connected to the FMCL to supply the core when the
outlet system is drained. Results obtained with interrupted natural circulation have enabled
us to formulate specifications for the forced cooling system operating during servicing with-
out boiling.
1. N. A. Dollezhal' and I'. Ya. Emel'yanov, The Channel Nuclear Power Reactor [in Russian],
Atomizdat, Moscow (1980).
2. V. N. Smolin et al.,.Aspects of Nuclear Science and Engineering, Reactor Design Series
[in Russian], Issue 1(8) (1974), p. 3.
GAS PHASE IN EXPERIMENTAL FUEL ELEMENTS WITH COMPACT
URANIUM DIOXIDE, IRRADIATED IN THE SM-2 REACTOR
A. P. Kirillovich, V. Sh. Sulaberidze,
Yu. I. Pimonov, V. N. Shulimov, Yu. G.
Lavrinovich, A. S. Biryukov, and
V. N. Kupriyanov
The intensive development of investigations of gas release from nuclear fuel is dictated
by practical problems of increasing the efficiency of fuel elements and the safety of their
subsequent reprocessing. A considerable amount of experimental data has been accumulated
already about the kinetics of escape of gaseous fission products from uranium dioxide [1,
2],
and the mechanism and development of numerical models of gas release were studied in [3,
4].
However, information about the chemical and isotopic compositions of the gas phase and
about the behavior of gaseous fission products in irradiated fuel elements with compact ura-
nium dioxide is very limited. This circumstance does not allow methods for calculating the
release of gaseous fission products from nuclear fuel to be verified by means of direct meas-
urements, and the amount of technological gases (oxygen, hydrogen, carbon compounds, etc.)
which can affect the efficiency of the fuel elements [1, 5], to be estimated. The results
of investigations of the amount and composition of the gaseous phase and the behavior of kryp-
ton, xenon, and helium in fuel elements with compacturanium dioxide and irradiated in-the
SM-2 reactor are presented below.
Translated from Atomnaya $nergiya, Vol. 57, No. 2, pp 91-95, August, 1984. Original
article submitted September 10, 1983.
0038-531X/84/5702-0517$08.50
? 1985 Plenum Publishing Corporation 517
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 Declassified and Approved For Release 2013/02/22 : CIA-RDP10-02196R000300050002-5
TABLE 1. Filling the Fuel Element,
System
and also the Residual Background of the
Sampling
d
Content, vol.
%
Gas being analyze
He
N2 + CO
02
Ar
CO2
Higqh- urity grade bottled
lle t'U-51-940-80)
99,999?5,9
0,00063?0,00012
0,00010?0,00002
0,000040?0,000008
0,000070?0,000014
Gas fillip fuel elements
99,93?5,9
0,040?0,008
0,01?0,002
0,0080?0,0016
0,0060?0,0012
Nos. 002 0023
Backkgroun o sampling
system:
before purging with He
99,96?5,9
0,20?0,004
0,010?0,002
0,009?0,002
0,00030?0,00002
after purging with He
99,98?5,9
0,010?0,002
0,0010?0,0002
0,00030?0,00005
0,00020?0,000(4
04
Fig. 1. Diagram of the gas sampler with the fuel-element punctur-
ing device: 1) fuel element; 2) device for puncturing irradiated
fuel elements; 3) filter; 4) PMT-4m lamp; 5) differential micro-
manometer; 6) helium bottle; 7) ampules for sampling the gas; 8)
vacuum pump.
Irradiation of the experimental fuel elements was carried out in a cell of.a channel of
the low-temperature water loop of the SM-2 research reactor. The design of the irradiation
facility, the pressure gauge in the fuel element, and the irradiation procedure were described
earlier in detail in [2]. The error in determining the linear power of the fuel element
amounted to ?7%, the average burnup ?8%, and the total error in determining the amount of gas
in the''fuel elements during irradiation did not exceed ?10%. When preparing fuel elements
0021, 0022, and 0023, the composition of the filling gas and the purity of the helium were
monitored by the mass-spectrometric method. The results of the analyses are given in Table 1.
Measurement of the Amount and Composition of the Gas Medium under the Jacket of the Ir-
radiated Fuel Elements. After cooling (from 5 to 15 months), measurements of the pressure
and sampling of the gas phase in the irradiated fuel elements were carried out by means of.
the sampling system (Fig. 1), located in a shielded chamber. The fuel element was placed in
a device for puncturing the can,. provided with a heater. The sampling system was hermetically
sealed and pumped out with ,a vacuum pump to a pressure of 1.6-2.6 Pa.
In the absence of inleakage during 30-40 min, the sampling system together with fuel
elements 0021, 0022, and 0023 was purged for 30 min with helium at a pressure of 40 kPa in
order to reduce the background from oxygen, nitrogen, and other gases. After pumping out the
system to 4 kPa, samples of the residual gas were taken and pumping out was continued to a
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 Declassified and Approved For Release 2013/02/22 CIA-RDP10-02196R000300050002-5
pressure of 1.6 to 2.6 Pa. Then the system-was filled again with helium up to 4 kPa, held
in this state for 30 min and, having sampled the residual gases (see Table 1), it was pumped
out to 1.6-2.6 Pa. The can was then punctured, and the gas released was sampled in the am-
pules. The amount of gas in the fuel element was determined by measuring the pressure (grade
of accuracy of the instrument 1.5) with a known volume of the system.
The investigation of the chemical and isotopic compositions of the gas in the irradiated
fuel elements was carried out by the mass-spectrometric method similar to that described in
[6]. An MI-1201 mass spectrometer was used for the measurements and was additionally provided
with a system for admitting the gas. The method of comparison with the sample to be analyzed
with a control gas mixture, prepared from certificated gases (i.e., additionally verified for
purity), was used.
The principal irradiation parameters and the characteristics of the experimental fuel
elements, taken from [2], are presented in Table 2. The fuel cores of the experimental fuel
elements, with a cladding of Kh18N1OT steel, were assembled from a plug of compacted uranium
dioxide enriched to 10% in 235U. The density of the fuel in fuel. element 001 was 10.43 ?
0.03 g/cm3, and in the remaining fuel elements it was 10.0 ? 0.1 g/cm3. The temperature of
the can during irradiation was 200?C.
The results of the determination of the quantity and composition of gas in the fuel ele-
ments after irradiation and cooling are given in Table 3. It can be seen that there is satis-
factory coincidence between the data on the quantity of gas under the fuel element can, ob-
tained during intrareactor measurements, and with puncturing of the can. The principal com-
.ponerits of the gaseous medium in the irradiated fuel elements (Table 3) are Xe, He, and Kr,
the volume content of which varies depending on the original quantity of filling gas and the
conditions of irradiation of the fuel.elements. In addition to gaseous fission products and
He in the irradiated fuel elements, Ar (0.001-0.06%), H2 ( 1 was satisfied in all the cases considered, the ratio
of the coefficients aT and aT corresponding to thermally stimulated desorption after ion and
electron bombardment proved not to be constant: It was much larger than 1 (curves 1 and 2 of
Fig. 3a and curves 1, I and 2,I of Fig. 3b), close to 1 (curves 3 and 4 of Fig. 3a), and was
even less than 1. As shown by the results of an experiment on cyclical bombardment of the
same copper target (Fig_, 4) the higher efficiency of thermally stimulated desorption after
electron bombardment (aT < aT) than after ion bombardment can be attributed to the higher
penetrating power of the electrons. In this experiment one cycle consists of cooling the
target to 78?K, bombarding it, and heating it to 293?K. In each of the first three cycles
the target was bombarded with 100-eV electrons (i = 80 pA) for 10 min (curves 1-3). Then it
was bombarded with ions (fourth cycle, curve 4) and electrons (fifth cycle, curve 5). The
irradiation parameters in the fourth and fifth cycles were the same: particle energy 6.7 keV,
beam current 0.015 pA; duration of bombardment 20 min. The decrease in aT with an increasing
number of cycles (curves 1-3, Fig. 4) as well as the decrease in a with a growing irradiation
dose (see Fig. 2) can be explained by depletion of the surface layer of the target in gaseous
impurities. The desorption rate should then depend on the impurity content. The ions pene-
trate to a considerably smaller depth into the metal than the electrons do and, therefore, the
energy absorbed per unit volume~of the metal is much higher during ion bombardment than during
electron bombardment. Estimates show [7] that 8-keV Mo ions and 200-eV electrons have rough-
ly the same effective depth range of_20 A (1 A = 10-10 m) in copper [6]. The energy of the
ions is converted mainly into energy exciting the electronic subsystem of the solid [4]. The
higher efficiency with which gaseous impurities are removed from the surface layer of the tar-
get under ion bombardment can be explained by the higher density of electronic excitations
(curve 4, Fig. 4). As a result of ion and electron bombardment the impurity distribution over
the surface layer of the target changes markedly. The surface layers are depleted in impuri-
ties as compared with deeper-lying layers. Then in the case of electron bombardment even a
comparatively low excitation density at a depth "unaffected" by ions can lead to a more pro-
nounced desorption effect (curve 5, Fig. 4) than under ion bombardment. The result of cycli-
cal bombardment (see Fig. 4) did not change when in one cycle during, the bombardment and some
time after the bombardment before the onset of heating of the target the residual hydrogen
pressure changed from roughly 10-8 to 1015 Pa, while the residual pressure of the gas with
M = 28 changed from 10-8 to 106 Pa. It can thus be concluded that the vacuum conditions do
not affect the stimulated purification of the target from gaseous impurities, i.e., the low-
temperature desorption was irreversible. In view of this the assumption is that the impurity
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Fig. 5. Dependence of desorption rate on the temperature dur-
ing annealing: 1-4) hydrogen desorption from copper after elec-
tron bombardment; 1) energy of bombarding electrons 500 eV, ie =
20 pA, duration of bombardment 6 min; 2) 4.5 keV, 0.005 pA, 40
min; 3) 5.8 keV, 0.015 pA, 20 min; 4) 9.2 keV, 0.02 pA, 15 min;
5) calculation.
center itself is an accumulator of activation energy and the process of stimulated desorption
is a method of relaxation of electronic excitations in a metal containing gaseous impurities.
From Fig. 4 we see that the maximum of the curve of thermally stimulated desorption after
bombardment of copper with electrons possessing an energy of a few kiloelectron volts (curve
5) is observed at a higher temperature Tm (the temperature of the maximum) than after ion bom-
bardment (curve 4). A similar displacement of Tm also appears for niobium (curves 3 and 4,
Fig. 3a). The tendency for the maximum of the curve of thermally stimulated desorption to be
displaced to a higher temperature as the electron energy is increased is clearly seen from
Fig. 5. As will be shown below the displacement of Tm is consistent with the concepts about
layer-by-layer removal of gaseous impurities from a target by ionizing radiation; The deeper
the bombarding particle penetrates into the target, the larger the volume that undergoes out-
gassing.
Volume diffusion of gaseous impurities when an electron-bombarded target heats up is al-
so indicated by the fact that after sufficiently prolonged low-temperature electron bombard-
ment the number of gas particles released is sufficient for several monoatomic layers. For ex-
ample, integration of curve 2 of Fig. 5, that was obtained when the copper target heated up
after 40-min bombardment with 4.5-keV electrons (i = 0.005 pA), indicates desorption of 3.1014
hydrogen molecules, i.e., 3.1015 cm 2. For thermally stimulated desorption of gas particles
it is necessary that the excited states of electrons, leading to stimulated diffusion, relax
after the particles reach the surface of the target. Hence, besides mass transfer stimulated
desorption should also be accompanied by a transfer of the energy of the electronic excitation.
Thus, radiation-stimulated desorption of gases from metals can be represented by the fol-
lowing scheme. When ionizing radiation acts on the metal, in addition to the formation of
complexes of different kinds [1, 2], there is excitation of some impurity centers which mi-
grate over the metal lattice, and reach the surface of the target, where the system gas parti-
cle-metal undergoes final decomposition.
If it is assumed that the process of thermally stimulated desorption is limited by the
volume diffusion of the excited impurity centers, in the absence of internal sinks the con-
centration of excited impurity centers in the target at a distance x from the surface at a
time t is given by the solution of the following boundary-value problem on the half-plane
(0