This item is only available as the following downloads:
Subject Category: PHYSICS
UNITED STATES ATOMIC ENERGY COMMISSION
A BRIE F DESCRIPTION OF A ONE
MEGAWATT CONVECTION COOLED
HOMOGENEOUS REACTOR -LAPRE II
L. D. P. King
April 13, 1955
Los Alamos Scientific Laboratory
University of California
Los Alamos, New Mexico
Technical Information Service, Oak Ridge, Tennessee
The Atomic Energy Commission makes no representation or warraonty
as to the accuracy or usefulness of the information or statements contain ned
in this report, or that the use of any information, apparatus, method or
process disclosed in this report may not infringe privately-owned rights.
The Commission assumes no Itabtlity with respect to the, ue of, or for
damages resulting from the use of, any 'Information, apparatus, method or
processt disclosed in this report.
Date D~eclassified: August 17l, 1955.
This report has been reproduced directly from the best
Reprfoduction of this information is encouraged by the
United States At~amic E~nergV CoHmmission~. Arrangement s for
your republicationa of this document in whole or in part
should be made with the author and the organization he
Issuance of this document does not constitute authority
for declassification of classified material of the same or
abmilar con~tent and title by the same autho~re.
Printed in USA, Price 20 cents. Available from the
Office of Tech~nical Services, Department of Commerce, Wash-
ington 25, D., c.
GPD BP2Ial I
A Brief Description of a One Megawatt Convection
Cooled Homogeneous Reactor -- Lapre II
By L. D. P. K~ing
This report has been prepared primarily to assist the Reactor Safeguard
Commilttee in evaluating LAPR2 II. This reactor is a simplified convection
version of LAPRE I which has been previously described for this Cormmittee in
LAM~S-1611. The location is completely outside of anyr building= and will only
make use of the existing~ facilities for handling contaminated liquids or gases.
Geogeraphie, meteorologic and seismic data are the same as for LAPRE I
and were fully described in LAMS-1611. No further material of this type hats
been included here.
Work performed under Contract No. W-7405-Eng-36.
I. The Design and Operation of the Reactor
A. General Purpose
It is proposed to construct a simplified convection version of LAPRE I
as soon as possible. The construction of such a unit is to help provide the L~os
Alamos Power Reactor Program with important information.
1. A long-term test of the corrosion and othr properties of uranium
phosphate solutions under actual operating conditions is dboolutely essential
before such solutions can have any practical value as a reactor fuel.
2. A one megawatt convection cooled hanageneous reactor using a
urranium and phosphate solution might well serve as a pilot model for a very
compact, simple and reliable1 portable reactor. IAPRE II: is to be operated
mo~re or less contiaosly for an extended period unless obvious faults develop.
There is to be a minto of instrumoentation and gadgetry. NTo operator is to be
to attendance except for startup or during changes to. running conditions.
B. Reactor Lrocation
The location of LAPRIE I in a cell of the main TA-j5 laboratory has
been fully described in LANB-1611. It is proposed to place LAPRE II outdoors in
a hole in the ground. Figure 1 above the general location relative to the
building and LAPRE I. The entire reactor core will be plce in the bottom of a
steel tank approximately 20 ft deep and 42 in, in diameter. The upper portion
of the tank vill be filled by a 9 ft concrete plug sealed to the tank. The top
of the tank and plug are at ground level.
C. Solution Properties
The solution chosen for LARE II is somewhat different from that
planned for LAPRE: I and described in considerable detail in LAEMS-1611. A
comparison of some of the properties of the two types of solutions is given in
the table below.
IAPRE I LAPRE II
Composition of soup 0.6i M U + 7.5 M BsP04 0.5 M U + 17*5 W BaSoe
Vapor pressure at b3000 5900 psi < 500 psi
Relative recoadbination rates
at 4j)000 1 25
Corrosion resistant materials
at 43000 PS, Au, graphite Pt, Au, graphite, Ag
Additional materials at 100oC Stainless steel Cu, stainless steel
J ------J !
1 B ,
Figure 2 illustrates relative liquid filling levels vs. temperature
for 9846 and 91.6L) ph~osphoric acid solutions. These bracket the proposed 95c5
solution for LAPRE II. An initial reactor filling of about 809 abould give an
approximate filling of 95) at the operating temperature of 4300C.
Figure 3 compares preliminary hydrogen and oxygen recombination rates
over solutions with low and high concentrations of phosphoric acid. Actual
radiolytic gas recodbinartion rates measured under reactor operating conditions,
for the low acid concentration case, have shown that the rates given in Fig. 5
are low due to the poor mixing conditions existing during the experiments. An
ove~rpressure of less than 20 psi is expected from radiolytic gas production for
the LAPRE II solution.
D. Reactor Specifications
Name LAPRE II
Location and operation TA-35, Los Alamos Scientific
Life test of corrosion protection
technique; demonstration of
safety and feasibility of convec-
tion cooled power package reactor.
Design in progress
0.8-1 My heat
U02 (=pop U235) dianolved in
RIsPO4, operated with 200 pai
Water and HaP04
Graphite (11 in.)
~ 15 in.
2. Reactor Vessel
Vapor region volume
Core region volume
Overall. length, bottom of vessel
to top of top flange
Inside diameter of vessel
Wall thickness of vessel
Beight of core region
Height of heat exchang~er region
Relative Volume Data
-- -- 0.34j M Von in 91.601 H3P04
--0.40 M U~a in 98.00 EsP04
Recambination Ratee ve. Temnperatuzw
for Typical Dilute and Concentrated
Phosphoric Acid Soups
U~a in 95.75 BsPOI
in 2.9 M1 5*70
t = time in hour
PG = prresure of radiolytic gae
Working pressure limit for vessel
Pressure and temperature in vessel
at yield point of steel
5. Fuel solution
Fuellesolution volume at operating
Fuel solution volume at room
Power density Ibased on core region
of 62.6 liters)
Fissionable material inventory
Nhximum: (0.4 M)
Ninimum: (0.2 M)
Vapor pressure of fuel solution at
operating temperature (45000)
Density of fuel solution (2500)
A:U in fuel solution
Radiolytic gas evolved
4. Heat Exchanger
Area (outside surface of tubing)
Coolant flow rate
Coolant pressure drop through tubes
Overall heat transfer coefficient
Method of fuel circulation
Puel solution circulation velocity
Stainless steel or A-536 F-22
steel (2-1/4q6 Crs 1& Nb) clad/
plated with Ag and Au.
Netal 0-ring plus seal weld
2300 pai, T50C0
~ 0.3 M 002 in 17.5 M (959) 5*704I
~ 7.70 kg--reflector shts out
~ 3.85 kg--reflector shis in
< 800 psi
208 for 0.5 M U0,
Adequartely removed by back
30.6 fta (for 0.8 Nw)
43 helical and 1 spiral coil;
21 ft of 5/16 in. O.D. x 1/8 in.
I.D. stainless steel, precious-
metal-clad tubing; per coil.
2180 Ib/hr at 0.8 EMr (4.36 g~alf/
70oF inlet, 6000F outlet
600 psi outlet
* 100 psi
Reactor coolant water to secondary
heat exchang;er to air radiator or
5. Reactor Control
Movable 6j in, thick annular graphite sleeve
around reactor vessel. Total calculated
reactivity effect of 11$. Maximum removal
rate pp = 10/see.
Excess pressure in reactor vessel returns
fuel to non-critical reservoir of 110 liter
If overpressure exceeds desired value, excess
is vented to waste disposal system through
duplicate relief valves.
Manual vent valve stops reactor by lowering
fuel into reservoir.
Puel input rate? controlled by (1) a flow-
10aiting orifice on pressurizing gas line,
(2) preset pressure on injection line to
reactor. Fuel input rate will be limited to
1200 ce/mlin or a n= 10 /sec. Fuel ejection
rate from reactor for a 300 pai overpressure
empties upper 12 in. of reactor in about
Z$ aec for a calculated decrease in reactiv-
ity of about 100.
Excess criticality for two
handled by burnable poison
Xenon override by solution
boron and lithium).
Earth fill (rhyolitic tuff p =- 1.4). Exclu-
sion area to insure 25 ft of earth between
reactor and personnel. Reservoir tank and
secondary heat exchanger also underground.
It is estimated that both gamma and neutron
dose rates will be below tolerance outside
the exclusion area with the reactor at full
power. Inside the fenced area the ogamma dose
rate could be several r/hr and this region
will therefore not be accessible until the
reactor power is reduced.
Fast avetragle 6.1 x 1018 n em2/see;
thermarl average 1.2 x 10'" n/cm"/sec.
This assumes 68)6 of LAPRE I fluxee.
E. General Description
The aim in designing this reactor was to make it as simple, reliable
and foolproof as possible.
Figure 4 shows a cut through the reactor proper. An overall layout of
components is indicated in Fig. 5-
Since the vapor pressure is low, a thin-walled vessel can be used for
the reactor, and solution transfer can be easily accomplished, even at full
operating temperatures, by gas pressures less than 1000 psi.
The solution is highly corrosive for most materials and satisfactory
pinhole-fre'e plated surfaces are difficult to achieve. Due to the simple vessel
shape and elimination of all internal vessel components except the heat exchanger,
cladding of the vessel looks quite feasible. Either a heavy 0.060 in, silver
cladding plated with 0.00) in. of gold or an all gold cladding approximately
0.010 in, thick will be used.
The reservoir tank which is cooled by a convection water loop will not
exceed 100oC temperatures and can therefore be made of copper or stainless steel.
The capacity of this tank is slightly greater than that of the reactor vessel so
that all the solution can be held without danger of losing fuel out the over-
pressure relief line.
The relief lines from the reservoir and enclosure tanks unn into the
contaminated disposal system available at TA-55.
The reflector consists of two concentric cylinders of graphite each
cut from a single piece.
The use of a secondary heat exchanger completely isolates the radio-
active portion of the reactor from the heat load. No instrumentation except for
thermoeouples and the primary feedwater pump exists in the reactor region. The
latter may be replaced by a steam jet pump. Power control is obtained by the
rate of coolant flow in the secondary loop.
Any neutron detectors or irradiation facilities will be placed in
vertical thimbles outside of the entire reactor unit. A single sealed vertical
rod will extend through the top shield plug to permit slow motion of the reflec-
tor shim sleeve. The outer tank will be gas tight and filled with nitrogen.
A small polonium-beryllium source placed under the reactor is used for
initial startup, thereafter the beryllium oxide block becomes the primary back-
ground neutron source.
- COPPER LINER
1" 1 1
0 I23 6 12
APPnRjx. SCAI.E IN INC HES
- 13 -
The entire reactor reflector and outer tank will be at approximately
reactor temperature, during operation due to the good insulation properties of
Hot and cold feedwater lines are well separated to reduce vessel
strains. The cold water flows into the central header and steam leaves at a
peripheral header near the flanges. The flange seal is removed from the high
An independent convection loop with air radiator is at all times
available to take care of the fission product heating if the solution is dis-
charged into the reservoir tank.
Initial startup will be carried out with the reflector shim about two-
thirds in and the secondary feedwanter loo; set for low power. The reactor and
reservoir tanks are pump'e~d out through the gas sampling tube (Fig. 3). One Los
Alamnos atmosphere (~ 590 mm Htg) of hydrogen is then admitted to the entire
Solution is then added to the reservoir tank through the filling tube.
The uranium concentration has previously been adjusted to the mnolarity calculated
to be sufficient for about two year operation with a burnable poison. Sufficient
solution is added to the tank so that when at operating temperature it would
fill the reactor to 95& and a few liters remain to cover the connecting pipe
The hydrogen tank with approximately 1000 lb maximum available pressure
is opened through a reducing valve system. As a safety precaution, a flov
limiting orifice is in the line whiich prevents the Lgas from forcing solution
into the reactor tank at a rate which would exceed a reactivity change of 10//0*0.
-Cold solution from the reservoir flows through the 1 em I.D. pipe into
the reactor at a rate controlled by the reducing valve pressure. 'd~hen sufficient
solution ha gone into the reactor to reach the cold critical volume the solution
begins to heat up at such a rate that its negative temperature coefficient e
(5.7 x 10-4/oC) uses up the excess reactivity produced by the further addition
Raeativity effects will be observed by the rise in solution temperature
as indicated by thermocouple A, Fig. 'j. One or more neutron chambers placed
outside the reactor outer tank will also be used during initial startups to
detect multiplication effects.
Thermocouple A vill continue to shlow an increase in temperature
throughout the solution addition period. When the reactor core proper is full,
convection will begin, and thermocouple F in the primary heat exchanger loop
will indicate a temperature rise. When the solution has reached the uppermost
cooling coil layer, thermocouple C in the steam header will indicate a sharp
rise. At this point the solution should be about at operating temperature. If
the temperature is slightly off, the reactivity will be corrected by the reflec-
tor shim. If there is a substantial difference between designl and actual temper-
ature, the uranium concentration of the solution can be changed.
reactor power can then be controlled by the throttle valve in the
secondary loop. Temperatures from thermocouples I and J and coolant flow from
the floymeter will permit output power calculation.
It is hoped that the reactor can be maintained at full operating
temperature and power for long periods of time. After initial testing no opera-
tor will be in attendance except for occasional routine checks.
II. Reactor Razards
A. Potential Kazards
Th following dangerous conitions might occur during operation of
1. Rupture of the reactor vessel.
2. Rupture of a heat exchaniger tube during operation.
3. Flooding of the reactor enclosure tan.
.4. Reactor power oscillations due to solution transfer between the
reactor vessel and reservoir.
5. Failure of a large portion of the vessel. protective cladding.
6. Rapid reactivity increases caused by solution injection, reflector
shim insertion, or full power demad from zero level.
7. Overheating of reservoir tank from fission product heat source.
8. Failure of coolant pump in primary or secondary loop.
B. Evaluation of hazards
1. An experiment using a scale model indicated that if the reactor
burst, the maximum pressure in the enclosure tank yould be one-half the reactor
pressure. An overpressure relief valve in the enclosure tank will take care of
such a rupture. This relief valve vill be set at about 2-i0 psi and will vent
into the vaste disposal system available at TA-~55
2. Simnulated rupture of a heat exchanger tube during operation was
tested in a scale model. Wlith the feeavater pressure exceeding the reactor
pressure by 200 pai, only a 70 pal pressure rise due to water injection vac
observed. Since the feedwater and vapor pressures will. probably differ by less
than 200 psi, no appreciable pressure rise should be produced.
3. Accidental flooding of the enclosuret tank during operation due to
a ruptured feedwater line could build up high pressures from the vaporization of
the wate~r. The relief valvet (see "2n) will. prevent rupture of the enclosure
4. Calculations have shown that appreciable power occillations cannot
be induced in the reactor vessel through its coiuplin; to the pressurized reser-
voir even for connecting pipes as large as 1 in. I.D. The choice of al e m I.D.
line is considered to offer no hazard in this respect.
5. Estimates based on corrosion data indicate that no large, rapid
pressure surges can occur from the exposure of large areas of pressure vessel
walls due to a failure of the cladding. A rate of about 0.2 poi/min/sq ft of
exposed surface vae dbtaine for stainless steel.
6. Calculations for maximum allowable reactivity variations indicate
that rapid changes should not exceed $1.00 in less than 1 sec. Reactivity
adjustment rates were arbitrarily limited to one-tenth of this, i.e., 10 /sec.
The maximum safe solution injection rate based on 10 l/see is 1200 ce/min. Rates
will be limited to values below this by a flow restricting orifice in the
pressurizing gas line.
The maximum safe reflector shim insertion rate, based on 10 /see,
is about 10 in./min. The hand operated lead screw will, be designed so that this
rate cannot be exceeded.
Calculations indicate that the peak power reached for a full
1.0 Mw power demand from a 10 watt level is about 11 Mw. The zero power level
for the reactor will normally be at least 1-2 kv to maintain temperature. The
maximum power overshoot for such an initial power will last for a few tenths of
a second and will not exceed 6 Mw. The fuel temperature rise will b~e only a few
degrees for such a burst.
7. If the entire volume of solution is discharged into the reservoir
tank. after long-period operation of the reactor, about 25 kvw of heat would be
produced in the reservoir from the fission products. An independent ceonvection
cooling loop and radiator system has been designed to hold the reservoir temper-
ature below 100oC in such a case. The vapor pressure of the solution will not,
therefore, rise to dangerous values when the fuel is dumped.
8. Failure of either the primary or secondary feedwater pumps will
cause the fuel temperature and vapor pressure to rise; the reactor will then
become subcritical and subsequent pressure and temperature rises will be at a
lower rate, since only the fission product heat source remains. The rising
pressure will dischrarge fuel into the reservoir until equilibrium is reached.
It is noted that most of the hazards listed are important because in
some fashion or other, they can produce an overpressure in the reactor vessel.
Setting the reservoir vent valves to release at the allowable working pressure
for the vessel (1500 psi) should give adequate protection. The fuel discharge
line appears to be large enough so that, once the reservoir vents, the pressure
in the reactor vessel cannot continue to rise.
Contributions to the material covered in this report have been obtained
from a large fraction of Groups K-1 and K-2. The following assisted in particular
in obtaining data and evaluating the hazards: E. 0. Swickcard, R. E. Peterson,
P. J. Bendt, R. M1. Kiehn, J. R. Phillips, R. M. Bidwell.
UNIVERSITY OF FLORIDA
3 1262 08917 0871
xml version 1.0 encoding UTF-8
REPORT xmlns http:www.fcla.edudlsmddaitss xmlns:xsi http:www.w3.org2001XMLSchema-instance xsi:schemaLocation http:www.fcla.edudlsmddaitssdaitssReport.xsd
INGEST IEID EFZ40CNAC_42C0R7 INGEST_TIME 2012-10-22T12:44:51Z PACKAGE AA00012263_00001
AGREEMENT_INFO ACCOUNT UF PROJECT UFDC