"OPERATING EXPERIENCE of the EBR

"OPERATING EXPERIENCE
CONF-830 905—12
of the
EBR-II INTERMEDIATE HEAT EXCHANGER
DE85 005040
and the
STEAM GENERATOR SYSTEM"
H. W. Buschman
K. J. Lonqua
W. H. Penney
ARGONNE NATIONAL. LABORATORY
EBR-II PROJECT
The submitted manuscript has bren authored
bv <T contractor of the U. S. Government
under contract
No. W-31-109-ENG-38.
Accordingly, the U. S. Government retains a
nonexclusive, royalty-free license to publish
or reproduce the published form of this
contribution, or allow others to do SD, for
U. S. Governmeni purposes.
To be Presented at the
ASME/IEEE
Joint Power Generation Conference
September 25-29, 1983
Indianapolis, Indiana
NOTICE
PMftQHS OF THIS REPORT ARE ItlEfflBU.
If has been reproduced from the bast
available copy to permit the broadest
possible availability.
*Work Supported by the U.S. Depan$me|lt of Energy and Electric Power Research
Institute
#• W
DISTRIBUTION OF THIS DQCUMfNT IS UNLIMITED
INTRODUCTION
Initial Purpose of EBR-II
Experimental Breeder Reactor-II (EBR-II) is an experimental liquid
metal fast breeder reactor located at the Idaho National Engineering
Laboratory. It consists of an unmods^ated, heterogeneous, sodium-cooled
reactor with a nominal thermal power output of 62.5 MW; an intermediate
closed loop of secondary sodium coolant; and a steam plant that produces 20
MW of electrical power through a coventional turbine generator.
EBR-II was originally designed as an engineering facility to demonstrate
the feasibility of fast reactors for central station power plant applications.
It was also intended to prove that a breeding ratio greater than unity could
be obtained in a power producing reactor. The EBR-II facility was also designed
to prove the feasibility of a completely integrated plant in which fuel could
be irradiated in the reactor, reprocessed in the Fuels and Examination
Facility, and returned to the reactor without being removed from the site.
The thermal performance of the reactor and the size of the system components
were intended to be amenable to direct extrapolation to central station
application. The plant was designed to permit a maximum of experimental
flexibility by separation of the plant systems, and yet permit extrapolation
to a commercial plant which would not require this same deqree of separation.
Evolution of EBR-II Mission
Experience with the reactor during early operation indicated that it would
also be useful as a high temperature fast neutron irradiation facility. Since
the long-range national emphasis had shifted to much larger plants using ceramic
fuel, the purpose of the facility was redirected in 1965 to provide irradiation
services for the development of fuels and structural materials for the Liquid
Metals Fast Breeder Reactor (LMFBR) program.
Recently, t!-- mission of EBR-II has expanded into a program of OperationalReliability Testing (ORT) which includes fuel-performance testing and operational
safety testing wherein the heat transport system and its components are required
to accommodate the consequences of transient power and duty-cycle events imposed
upon advanced fuel-element designs.
-2Generai Layout of EBR-II
The EBR-II reactor is a pool-type design where all primary system components are located in a large sodium filled tank. The general arrangement of
the system is shov.n in Fig. 1. The primary pumps are in the cold leg piping
and take their suction from the pool. The flow is directed through the reactor,
where it is heated by nuclear fission. From the reactor, the hot sodium flows
to the IHX and then returns to the pool. In the IHX, heat from the radioactive
primary system is transferred to the secondary sodium. The secondary sodium
system is essentially nonradioactive and is used to transfer heat from the
radioactive primary system located inside a containment buTdiig to the
steam generating equipment located outside the containment ouilding.
The EBR-II steam generator system uses heat from the reactor by way of
the primary and secondary liquid-sodium cooling systems. The steam generator
system consists of seven (formerly eight) natural circulation evaporators,
two once-through superheaters, and a single steam drum. The steam drum is
located horizontally above the evaporators and superheaters. The evaporators
are arranged in two rows and are connected to the steam drum by individual
risers and downcomers. Primary and secondary steam separation takes place
within the drum and dry-saturated steam is routed from the top of the drum,
through the parallel connected superheaters to a common header to the turbine.
Feedwater is supplied to the drum, where it mixes with the saturated steamwater mixture before entering the downcomers. Blowdown is taken from a
collection header located within and near the bottom of the drum. Figures 2
and 3 are schematic of the general layout of the system.
DISCLAIMER
This report was prepared as an account of work sponsored by an agency of the United Slates
Government. Neither the United Stales Government nor any agency thereof, nor any of their
employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or
process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark,
manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views
and opinions of authors expressed herein do not necessarily state or reflect those of the
United States Government or any agency thereof.
. -3PERFORMANCE OF THE EBR-II HEAT TRANSPORT SYSTEM
Operating Parameters (See Table 1)
Secondary sodium flows from the intermediate heat exchanger (IHX) located
in the primary-sodium tank to the steam generator at about 299.5 kg/s
(5400 gpm) and 466°C (870°F), then passes through the superheaters, where it
raises 30.9 kg/s (245,000 Ib/hr) of 304°C (580°F) saturated steam to 438°C
(820°F) superheated steam. The secondary sodium then passes on to the
evaporator inlet headers at about 427°C (800°F), and then passes through the
evaporators, heating steam-drum water in an almost isothermal process to 304°C
(580°F) wet steam, which returns to the drum. The sodium then returns at
about 307°C (585°F) to the IHX.
Feedwater is normally supplied to the steam drum at 33.4 kg/s (265,000
lb/hr) and 288°C (550°F); this accounts for a continuous blowdown flow of
2.5 kg/s (20,000 lb/hr), which is extracted from the steam drum. To inhibit
corrosion of the systems, the feedwater is chemically treated by injection of
hydrazine and morpholine. Hydrazine (N?H«) is used to scavenge dissolved
oxygen in the feedwater. Morpholine (C.HgNO) is used to maintain feedwater
pH in the range of 8.6 to 9.4. In addition, the morpho'line acts as a
neutralizing agent for dissolved CO^ or other acid-forming compounds. The
results of the feedwater treatment are shown in Table 2.
Operating History
The EBR-II heat transport system continues to operate satisfactorily
after 18 years. This represents about 89,000 hours of steaming, which
results in a total integrated thermal power production of about 215,000 MWd.
In this time, the steam generator has experienced over 580 plant startups and
349 reactor scrams. The plant capacity factor for the past five years has
been in excess of 70%, and in fact has averaged almost 60% over the last thirteen
years. This excellent record is partly attributable to the trouble-free
operation of the steam generator which, aside from an initial construction
tube-to-tubesheet weld defect, has had a plant availability of 100%.
-4-
THE INTERMEDIATE HEAT EXCHANGER
Configuration
The IHX (see Fig. 4) consists of three basic structures:
casing, (2) Tube bundle, and (3) Shield plug.
(1) Well
The weir casing is a cylindrical Type 304 stainless steel structure,
approximately 18.5 ft (5.64 m) long and 6 ft (1.83 m) in diameter. This
structure is an extension of the heat-exchanger nozzle of the primary-tank
cover. It provides the support structure for the primary-flow inlet diffuser
and neutron shielding that surround the heat-exchanger tube bundle. The tube
bundle and shield plug form an integral unit that slides into the well casing
from the top of the primary tank.
If heat-exchanger maintenance is ever necessary, the tube bundle and
shield plug can be removed from the well casing. Removal is accomplished
by draining the secondary sodium, cutting the secondary inlet and outlet
piping, breaking the upper mounting flange, and lifting the tube bundle and
shield plug out of the well casing. Since an inert-gas blanket must be
maintained at all times, a caisson or similar mechanism must be used during
the removal procedure. After removal, the tank nozzle must be closed with
a temporary plug.
The heat exchanger was designed with a low length-to-diameter ratio,
which is compatible with the philosophy of a low-pressure-drop heat exchanger.
The pressure drop was further reduced by maintaining axial flow to the maximum
practical extent. No provisions were made for cross flow, and the supportbaffle flow areas were maximized by the use of convoluted-ribbon-type supports,
rather than the more conventional drilled-plate-type support. A maximumpressure-drop criterion of 5 psi (34.4 kPa) for both primary and secondary
coolant was easily achieved. Values at full power operation are approximately
2.1 psi (14.5 kPa) and 3.5 psi (24.1 kPa) respectively.
Because of an unusually low length-to-diameter ratio for the heat
exchanger, a situation was created in which the primary flow could readily
become imbalanced. With imbalanced flow, much of the primary sodium would
-5not penetrate the tube bundle and would therefore bypass the center tubes.
This situation could cause a loss of overall efficiency and produce excessive
thermal stresses in the tubes and tube-to-tubesheet welds. To achieve balanced
primary flow, the heat exchanger was designed to provide an equal static
pressure drop, with proper flow for every possible flow path. Since good
thermal-convection characteristics were a requirement, cross-flow baffles were
considered to be unacceptable for use in the heat exchanger. As an alternative
to baffles, the belt diffuser and two orifice plates were used to achieve
equal pressure drops for all possible flow paths.
The secondary side of the heat exchanger was also required to have balanced
flow and good thermal-convection characteristics. The physical arrangement
of the secondary side was also designed to promote natural convection flow.
The secondary sodium enters the heat exchanger through an insulated pipe and
flows down to the lower ellipsoidal head. Within the head, the flow must
make a 180-degree turn before flowing up through the tubes. A semi-torusshaped diffuser is enclosed within the lower head to turn the flow the required
180 degrees; the diffuser also distributes the coolant to provide a balanced
secondary flow.
IHX Performance
Thirty-four thermocouples were installed to provide temperature data at
various locations on the primary-sodium side (shell side) of the heat exchanger.
Eight of these thermocouples were installed at various locations just below the
top orivice plate. Eighteen were at various locations below the bottom
orifice plate. Four each were positioned to monitor the primary-sodium inlet
and outlet temperatures.
These thermocouples provided little useful information because most
of them had failed early in life (prior to raising power above 45 MW). Data
apparently were not systematically recorded and/or reported early in life from
these thermocouples, so perfomance data are available only from indirect plant
data. No instrumentation was provided to directly measure secondary-sodium
temperatures, or pressures, or pressure drops within the assembly.
-6Based upon plant instrumentation external to the IHX, the overall heattransfer performance of the EBR-II IHX has been in agreement with the design
correlation. Considering the unknown channeling of hot sodium on the inside
and outside of the unit, which would lower the overall performance, the
system condition was adequately described by the 'design correlations
initially used.
Temperature measurements from, the installed instrumentation indicated
that some of the primary sodium is short-circuiting the tube bundle and
traversing the unit essentially uncooled. This occurs in the open areas
in the tube bundle next to the center pipe and at the outer periphery
next to the shell. This uncooled sodium is not forced to mix with the cooler
sodium until the flow streams reach the lower orifice plate. Temperatures
have been measured near the inner and outer peripheries of the tube bundle
below the iower orifice plate, after some mixing has occurred; these
temperatures are of the order of 820°F (438°C). This compares with an
average outlet temperature cf 700°F (371°C). This measurement was made
at full power when the hot primary inlet temperature was 883°F (473°C).
This bypass flow lowers the performance of the exchanger and would explain
why the measured performance is less than design, as previously discussed.
One other observation, which has caused some minor operational concern
but has not resulted in any real problem, is worth mentioning. When primary
flow is established through the tube bundle, the pressure in the belt diffuser
is equal to the pressure drop through the shell side. This causes primary
sodium to rise in the annulus between the shield plug and nozzle casing. With
a pressure drop of 2.1 psi (14.48 kPa), the sodium rises as much as 5 ft
(1.52 m) up the annulus. With flow changes, this causes a washing action in this
annulus as the level moves up and down. As a result, higher than normal
temperature and radiation levels have been observed in and above the primarytank cover in the vicinity of the IHX. Another concern is the thermal stress
cycling that occurs as a result of this washing action at the weld joining the
1-in.-thick (25.4-mm) well casing to the 2-in.-thick (50.8-mm) bottom plate
of the reactor-tank cover.
-7IHX Maintenance Events
Except for a minor problem in November 1970, when it was necessary to
remove the permanently installed evacuation tube, service has been troublefree. The investigation of the noise caused by the evacuation tube and
the activities involved in the repair are reported in detail in Reference (1).
The abstract from this reference adequately describes, for the purpose of
this presentation, the problem and subsequent repair.
"On the night of November 14, 1970, a loud banging noise was
heard in the vicinity of the EBR-II Intermediate Heat
Exchanger (IHX). Indications were that the noise source was
within the IHX inlet pipe. A port for access to the IHX
internals was installed on the inlet-pipe elbow. Visual
examinations using both a periscope and a remote TV system
revealed that of the two supports clips holding a 1-in.
(25.4-mm) diameter evacuation tube in place, the top clip was
loose and the bottom clip was missing. This condition allowed
the evacuction tube to move because of the secondary sodium
flow stream and vibrate against the wall of the 12-in. (3.24mm O.D) diameter inlet pipe. Evidence of wear on both the
12-inch (324-mm O.D.) pipe and the 1-in. (25.4-mm) tube was
found.
The upper clip was removed; the evacuation tube was cut at the
top and bottom and removed. The lower clip was not found.
The section cut out of the inlet elbow was rewelded in place
and the secondary system was restored to operational status.
Quiet operation of the IHX verified that the repair was
successful."
THE STEAM GENERATORS (EVAPORATORS)
Configuration
The evaporators are shell-and-tube-type heat exchangers. The basic
material used in fabrication is 2-1/4 Cr - 1 Mo ferritic steel. This material
was selected because of its favorable high temperature properties, established
use (ASME coded), availability, cost, fabricability, and resistance to
caustic attack. Baffle plates are Type 304 stainless steel. The evaporators
are designed to minimize the possibility of interaction between sodium and
water/steam by using bonded duplex tubes and double tubesheets (see Figs. 5,
6, and 7 ) . One tubesheet seals the outer of the duplex tubes to form the
sodium cavity, and the other seals the inner of the duplex tubes to the
water/steam header. Two types of duplex tubing were used in the fabrication
of the units. Four evaporators contain mechanically bonded duplex tubes,
and four units contain metallurgically bonded tubes. Fabrication of both
types of tubes consisted of placing the outer tube over the inner tube,
drawing the two tubes together through a die and over a pin. This was
followed by expanding the duplex tubes by drawing a pin through the I.D.
without restraining the O.D. Metallurgical bonding required a final operation
of heating to flow the nickel-nickel phosphorous alloy between the tubes to
produce a brazed tube-to-tube-bond. The heating operation annealed out the
prestress, which was introduced during the drawing operation, the mechanically
bonded tubes were left in the stressed condition of outer tubes in tension
and inner tubes in compression.
Maintenance Events
Initial power of EBR-II was attained in July 1964. On February 7, 1965,
during a shutdown period with the steam system at ambient temperature, the
operating crew reported that liquid water could be observed in the space
between the steam and sodium tubesheets at the upper end of evaporator .702.
The source of water was traced to a crater crack in one of the tube-to-tubesheet welds. It was obviously a birth defect that was not detected by the
original helium leak test because it was at least partially plugged with slag.
The defect was repaired by manual welding. Access was gained simply by
removing a section of the steam riser from the evaporator. Up to the present
-9time, no additional leaks have been detected on any of the steam generators.
The evaporator that experienced a leak in 1965 has been subjected to
periodic inspections of the steam side. The section of the riser that
was removed for leak repair was reinstalled with removable flanges to permit
reasonably easy access. Inspections were made in 1969, 1970, 1972, 1973, 1974,
1976, and 1978. Until 1974 all inspections were visual. Tube internals
were viewed with a borescope, and photographs were taken for comparison with
previous examinations. Surfaces were found to be covered with brown-ied
magnetic-iron oxide varying in thickness from light to 1.59 mm (1/16- n.).
The outer layer was a light porous coating, easily removed by brushing; the
inner deposit was a more adherent, dense film. There was no evidence of
pitting or metal loss when the deposits were removed. After the evaporator
tubes were brushed to remove the light deposits, axial and circumferential
fabrication marks could be seen through the borescope. The top 381 mm
(15-in.) of the evaporator tube was covered with a thin, fine-grained
deposit of iron oxide. The next 2.44-2.74 m (8-9 ft) from the top appeared
crystalline as seen through the borescope. The crystalline appearance
disappeared gradually down the tube length. Samples of the deposits on the
tubesheet have been analyzed and found to be primarily iron, copper, and
nickel. Starting in 1974, ultrasonic techniques have been employed to examine
the duplex tubes of EV-702. These tubes are metallurgically bonded. The
techniques and results of these examinations are reported in Reference (2).
The results of these examinations .ndicate a complete lack of any indications
that are detrimental to the integrity of the evaporator.
In 1980, evaporator EV-706 was removed from the system and coverted to
a superheater for later replacement of one of the superheaters. Ultrasonic
inspection of all the steam tubes revealed no defects. The steam tube-totubesheet welds were examined using the liquid penetrant method. The tube
I.D.'s were measured and found to be within manufacturing tolerance. Except
for tht same deposits observed in inspections of evaporator EV-702 and
described above, the unit was in excellent condition and considered acceptable
for conversion to and installation as a superheater.
-10THE SUPERHEATERS
Performance
In 1974 the superheater containing mechanically bonded tubes began
exhibiting anomalous thermal behavior. The anomalous behavior is typified
by a sudden decrease in outlet steam temperature occurring just as full power
is approached (Fig. 8 ) . The magnitude of the decrease in outlet temperature
varies from startup to startup and has not been correlated with any specific
plant parameters. Detailed analyses and examinations are in process to better
define the mechanisms contributing to this behavior. The magnitude of the
decrease generally increased with time, and the power level at which the drop
occurs seems to be decreasing. This trend is shown in Figure 9, which shows the
difference in steam outlet temperatures between the two superheaters as a
function of calendar time. Measurments have shown that individual tubes.are
exhibiting sudden drops in steam temperature (Fig. 10) and that the magnitude
of the average outlet temperature is dependent on the number of individual tubes
so performing. This behavior is explained by an increased thermal resistance of
the duplex tubes caused by a reduction in the contact pressure, and in some
cases, actual separation. Tube separation occurs when the differential thermal
expansion between the inner and outer tube is sufficient to overcome the
effect of residual tube prestress combined with steam system pressure within
the inner tube. The increase in thermal resistance caused by reduced contact
pressure or tube separation results in a larger temperature difference between
the two tubes, which leads to further separation. This process continues,
along with a decrease in the superheater power level, until stable heat transfer
conditions exist.
Maintenance Events
An in-service inspection or the superheater containing mechanically bonded
duplex tubes, SU-712, was made in April 1979. This examination was prompted
by the observed anomalous thermal behavior reported herein. The main activities
of the inspection included visual and ultrasonic inspection of the steam-side
internals and the removal and destructive examination of three of the core
tubes. Straightness and I.D. measurements of three steam tubes were made.
-11Ultrasonic inspection of these three tubes indicated that the inner and outer
tubes were in intimate contact. Although many interesting observations and
measurements were made, the overall assessment of the examination results is
that the unit is surprisingly clean and appeared to be "like new" and that
there are no unusual conditions or any unusual wear. The details of the
techniques used and the results are reported in Reference (3).
Superheater SU712 Disassembly and Examination
In April 1981, the superheater was removed from the EBR-II steam system
for destructive examination. The sodium side was sealed and maintained in an
inert argon atmosphere until it was cleaned of residual sodium. The steam
side was dried and sealed until examination was started in June 1981. Details
of the disassembly examinations are found in Reference 4.
The sodium side was cleaned with ethanol prior to disassembly. After
cleaning, the exposed surfaces were free of sodium; but restricted interfaces retained unreacted sodium along with some reaction products.
The superheater was disassembled in a sequence that would progressively yield examination results prior to their obliteration by subsequent
disassembly. The inlet and outlet steam reducers were first removed (Fig. 11)
to allow access to the steam tube inside surfaces for visual examination, 1.0.
measurements, and straightness measurements. The steam surfaces contained
lightly scattered corrosion pitting. The pits were less than 0.25 mm (.010-in.)
in depth and appeared to have been formed early in the life of the tubing as
evidenced by the oxide coating. The inside diameter measurements were within
0.05 mm (.002-in.) of the nominal fabricated diameter of 27.05 mm (1.065-in.).
The results of the straightness measurements indicated that some tubes were
bowed beyond the maximum measurable offset of 20.6 mm (0.810-in.), that the
peripheral tubes were bowed more than the central tubes, and that the bowing
was consistently in the same direction, i.e., the tube bundle was twisted in one
direction then returned in the opposite direction (Fig. 12)..
-12Because the superheater was fabricated with a prestress imposed on the
shell and tubes (shell in tension and tubes in compression) the disassembly
was conducted in a manner that permitted measurement of the residual prestress
by two methods. Strain gages were placed on the shell and on the outside
surface of one accessible tube. Benchmarks were applied to the shell on
either side of a lateral cutting plane. The shell was cut and the springback,
and change of shell strain, were measurad. The measured springback and the
strain gage results closely correlated. The final prestress was 55% of the
design prestress.
The tube-to-tubesheet welds were a matter of concern since they were at
the location of highest stress. Liquid Penetrant examination located
discontinuities in some weld craters (Fig. 13). These discontinuities are
being evaluated by metallographic examination.
The superheater was essentially an evaporator installed in an
inverted position. The baffle nest, which thus was suspended from inadequate
welds, had broken away from its support ring. It was found displaced about
22 inches, but had "fl jated" in the flow stream during operation within 2-in.
of its as-built location (Fig. 14).
In general, the superheater was found to be in remarkably "like new"
condition with even the original soap stone marks clearly visible on the
baffle nest.
Materials Examinations
Materials examinations of the duplex tubing are in process; some
results are reported below. Examinations of the remainder of the subcomponents,
i.e., the shell, the baffle nest, the tubsheets, and the inlet and outlet
reducers revealed no deleterious effects of operation.
-13Duplex Tubing Examinations
To determine the contact pressure at the interface (radial stress)
between the tubes, the Sachs method was used. This method requires the
installation of strain gauges on either the tube bore or outside surface.
These strain gauges measure the change in circumferential and longitudinal
strain as either the bore or outside" surfaces are removed in small increments.
The use of this method allows the calculation of radial stress (contact pressure)
tangential stress, and longitudinal stress. The testing program required
specimens from three different archive tubes and four superheater tubes. Two
of those tubes were known to have reduced steam temperature. One had normal
steam temperature, and one was both reduced and normal at various times in
life. Test specimens, 153 mm long, were cut from four locations within the
tube; near the steam inlet, near the steam outlet, the midpoint, and 7880 mm
from the steam inlet. Additional specimens (153 mm) were cut from the
three archive tubes. Two strain gauges were mounted in the circumferential
direction, 180° apart, and two strain gauges in the longitudinal direction,
180° apart. The specimen was placed in a fixture and 0.25 mm increments
were bored from the inside with measurements obtained after each increment.
The residual stress (radial, tangential, and longitudinal) distributions
were then calculated by means of the Sachs equations.
The contact pressures obtained from the tubes in the superheater were
lower than those obtained from the archive tubes. Also the contact pressures
were highest at the steam inlet/sodium outlet end of the unit, the end with
the lowest operating temperature (Fig. 15). The steam outlet/sodium inlet
had the lowest contact pressure, and was the end with the highest operating
temperature. It appears the contact pressures were temperature sensitive.
Thus, it is feasible that some creep or stress relaxation of the material has
occurred in the tubes. All of the tubes examined reveal the same trend in
contact pressure relaxation.
Tensile test specimens were taken from the tube wall adjoining the
Sachs specimens. The results of these tests, indicative of a stress
relaxation anneal, confirm the trends obtained from the Sachs specimen.
-14The tensile test specimens indicated a reduction in ultimate tensile strength
and increase in elongation as the distance from the steam inlet/sodium outlet
increased.
-15ACKNOWLEPGEMENTS
The work reported herein was performed under the auspices of the United
States Department of Energy with the support of the Electric Power Research
Institute.
REFERENCES
(1) Buschman, H. W., Cerutti, B. C , and Clark, A. F., Noise Investigation
and Repair of the E3R-II Intermediate Heat Exchanger," Argonne National
Laboratory report ANL-7834, August 1971.
(2) K. J. Longua, et al, Second Mechanical In-Service Inspection of E1R-II
Steam Generator, Argonne National Laboratory report ANL-78-86~ November
1978.
(3) Penney, Wm. H.s et al, In-Service Inspection of Superheater 712 at
EBR-II, Argonne National Laboratory report ANL-80-2, March 1980.
(4) Penney, Wm. H., Buschman, H. W., and Washburn, R. A., Disassembly
and Phase 1 Examination of EBR-II Superheater SU-712, Argonne National
Laboratory report ANL-82-16, Hay 1982.
TABLE 1
EBR-I.I HEAT TRANSPORT SYSTEM OPERATING PARAMETERS
Power
Secondary Sodium Flow
Secondary Sodium Temp.
From IHX
To Evap.
To IHX
Feedwater Temperature
Steam Drum Pressure
62.5 MWt
299.5 kg/s (5400 gpm)
466°C (87O°F)
427°C (800°F)
307°C (585°F)
288°C (550°F)
9.14 MPa (T t = 304°C)
(1325 Psi, Tsat = 580°F)
Superheated Steam
Temperature
Blowdown
Steam Flow
Circulation Ratio(c)
438°C (820°F)
2.5 kg/s (20,000 lb/h)
30.9 kg/s (245,000 Ib/h)
10
TABLE 2
FEEDWATER QUALITY DURING POWER OPERATION
EBR-II a
LMFBR
,
Conceptual
Design
Study
Total Dissolved Solids
%o
50 ppb (max.)
Dissolved Oxygen
< 5 ppb
5 ppb (max.)
Silica
<10 ppb
10 ppb (max.)
Iron
<10 ppb
10 ppb (max.)
Copper
<2O ppb
PH
8.6 - 9.2
8.5 - 9.3
10 - 20 ppb
2 0 - 1 0 0 ppb
Sodi um
<0.1 ppb
< 3 ppb
Chlorides
< 20 ppb
< 3 ppb
Hydrazine
a
(Residual)
< 2 ppb
Concentrations shown as < are lower than measurement capability.
Unpublished data
Figure 1
FEEDWATER
280°C (55O°F)
33.4 kg/s {265,000 lb/hr)
BLOWOOWN
304°C (580°F) _,
2.5 kg/s (20,000 lb/hrf
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Figure 3
STEAM DRUM
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TURBINE'
RS(8)
SUPERHEATERS (2)
DIRECTION OF
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F iguie 4
PIPE SIOP
LIFTING LUG
HEAT EXCHANGER
(U) NOZZLE
SECONDARY SODIUM INLET
' SECONDARY SODIUM OUTLET
CENTRAL INLET PIPE -
PIPE SHROUD •
PRIMARY TANK COVER
ARR(1N FILLED ANNULAR SPACE
THERMAL INSULATION
•1UBES
. IMPACT BAFFLES
UPPER TUBESHEET •
- 8f.LT DIFFUSER
UPPER ORIFICE PLATE •
GUIDE BARSPRIMARY SODIUM INI ET
ANNULAR SPACE"DIAPHRAGM SEALS
TYPICAL TUBE-
1.5 W"i BORON SST
NEUTRON SHIELD
SUPPORT SLATS
" ^ ^ V PRIMARY SODIUM OUTLET
LOWER TUBESHEET AND THERMAL BARRIER
-LOWER HEAD
SEMI-TORUS FLOW DIFF.USER
-INSTALLATION GUIDES
BORAL PLATE (INSiP
DOUBLE • WALLED HEAD)
• EVACUATION TUBE
RE-6-49072-C
Intermediale Heat.Exchanger
Figure 5
SATURATED
STEAM INLET
STEAM TUBESHEET
SODIUM TUBESHEET
S l f l l K A n D SHAM OUTLfT
IIW'F)
Figure 6
INSPECTION ACCESS OPENING
STEAM/WATER
SIDE
TO TUBE
SHEET WELD
.SINGLE WALL
TUBE
TUBE SHEET
TUBE TO TUBESHEET WELD
Figure 7
WATER/STEAM HEADER
Steam
Tubesheet
A1r
Space
Sodium
Tubesheet
Sodium
Cavity
Duplex
Tube
TUBE DIAMETERS
Inner Tube
Outer Tube
I.D. - 27.05 BW
(1.065 I n . )
I . D . - 3 1 . 7 5 mm
(1.250 1n.)
O.D. - 3 1 . 7 5 mm
(1.250 i n . )
O.D. - 3 6 . 5 3 mm
(1.438 i n . )
187 6 754 AUG 798.3
858
MIN 764.S
STD 13.85
1
MAX 888.9
838.9
L
819.8
r
F
790.0
778.8
»
I
750.0
HRS.
1
21-04 MAY 16,'79 TO 23^04 MAY 16,'79
18B SUPERHEATER 718 STEAM OUTLET TEMP
2 HOUR LOG
EBR-II DAS
\
(2/81)
REACTOR POWER - 62.5 MW
50
•
9
o
LU
'_>
9
UJ
'9
9
9
30
to
c
CD
LU
9
9
20
9
LU
9
••
•
9
•
••
9V
LU
#
10
99
0
tr—m • J
9
J
9%
W
1/72
9 »
1/73
1/71
1/75
1/77
1/78
1/76
DATE
SUPERHEATER STEAM OUTLET TEMPERATURE DIFFERENCE
1/79
1/80
Figure 10
CM
CM
U.
e i
..-,:. if*
Figure 14
20
I
-20
CO
-40
LU
>
LLJ
i
r
o
0
CO
CO
LU
i
o-
\
-60
o
^
/ DISTANCE FROM
STEAM INLET, mm
549
3840
6152
7800
-80
X
o
X
-100
I
13
I
I
I
14
J
I
L I I V I
I I
15
16
RADIUS r
1
17
mm
1 I I I1
18
it
I
19
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