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AE-269UDC 621.039.55:621.039.51

621.039.524.46.634 (Agesta)

Experimental Equipment for Physics Studies

in the Agesta Reactor

G.Bernander, P. E. Blomberg and P.-Q. Dubois

AKTIEBOLAGET ATOMENERGI

STOCKHOLM, SWEDEN 1967

AE-269

EXPERIMENTAL EQUIPMENT FOR PHYSICS STUDIES

IN THE AGESTA REACTOR

G Bernander, ASEA Nuclear Power Department,

P E Blomberg and P O Dubois, AB Atomenergi, Sweden

SUMMARY

Comprehensive physics measurements were carried out in

connection with the start up of the Agesta reactor. For this purpose

special experimental equipment was constructed and installed in the

reactor. Parts of this were indispensable and/or time-saving for the

reactivity control during the core build-up period and during the first

criticality studies.

This report gives mainly a detailed description of the experi-

mental equipment used, but also the relevant physics background and

the experience gained during the performance.

Printed and distributed in March 1967

LIST OF CONTENTS

1. INTRODUCTION 3

2. DESCRIPTION OF THE PLANT 3

3. SCOPE OF EXPERIMENTS 5

3. 1 Buckling measurements 5

3.2 Reactivity coefficients 6

3.3 Control rod worths 7

3.4 Form factors 7

3.5 Microscopic studies 8

3.6 Dynamic studies 8

3.7 Fuel burnup 9

4. EXPERIMENTAL EQUIPMENT 10

4.1 In-core detectors 10

4.2 D 2 ° l e v e l m e t e r 13

4.3 Pulsed neutron source 15

4.4 Activation wire holders 16

4. 5 Special fuel assembly for fine structure studies 18

4.6 Equipment for burnup studies 19

4.7 Reactivity oscillator 23

4.8 The data acquisition unit RAMSES 24

REFERENCES 26

-- 3 ~

1. INTRODUCTION

When planning physics experiments in a power reactor core,

one is limited by the conditions resulting from the specific reactor de-

sign, especially when experimental features have not been built into ic.

The preparation of experimental equipment will then be a matter of

adaptation to existing conditions and is concerned primarily with the

question of accessibility to the core as well as the operability of the

reactor at somewhat unusual or non-ordinary states. We believe it is

of interest to demonstrate the extent to which a power reactor of the

Agesta type F 1, 2, 3] may be utilized to yield important reactor neutron

physics data. Although this is our main objective, we wish first to pro-

vide the experimental background, giving a summary of the types of

measurements and methods employed, at the same time referring to

some significant experimental results.

2. DESCRIPTION OF THE PLANT

The Âgesta Nuclear Power Station outside Stockholm, Sweden,

has a natural uranium, heavy water moderated, reflected and cooled

reactor f 1,2,3], Rated at 65 MWth (stage 1), it produces 10 MW of

electricity while the remaining power is delivered for space heating pur-

poses to a Stockholm suburban area. Being a first generation prototype

in the development of Swedish heavy water moderated power reactors,

the reactor has been subjected to extensive technological and physical

studies.

Criticality was initially achieved in July, 1963. A period of

complementary system and component tests and continued fuel loading

followed together with the low power physics experiments. High power

operation was begun in March, 1964. The power was then raised

gradually to 100 % in the course of a few days, most of which time

was devoted to physics measurements and assessment of control pro-

pe rtie s.

The flow diagram, Fig. 1, shows schematically the main

features of the reactor system. The reactor is of the pressure vessel

type and operates at a pressure of 33 bar. Heavy water as coolant is

• 4 -

Pressurizer Exit H2O steam toturbine (198°C.15 bar)

UO fuel

D2Omoderator220 8C

205 °C

Reactorpressurevessel

Steam^generator

Feed water ("K0

Four primary main circuits

Fig. I Flow diagram of Agesta reactor.

circulated at the rate of 1200 kg/s. From a bottom plenum the coolant

is distributed to the 140 fuel coolant channels, flows into the moderator

space and is then discharged at 220 C to the four steam generators.

The exit light water steam from these flows to a 12 MW back pressure

turbine and the associated condenser is used as a heat exchanger for

the district heating system. Excess heat may be disposed of using two

auxiliary steam condensers and a cooling tower. Prior to power opera-

tion the reactor is heated to 215 C by an electric heater in the second-

ary circuit feed water system, thus reversing the steam generators.

The UO? fuel assembly comprises four 19-rod bundles con-

tained in a coolant tube, the structural material being Zircaloy 2. The

37 loading channels, of which about half normally house the hydraulic

control rod drive units, each give access to four fuel assembly posi-

tions (in a square array) adjacent to the channel. The tubular Cd-In-Ag

control rods therefore move on the cell boundaries of the regular lattice

- 5 -

cells. The loading channels or standpipes thus permit access to the

core without any fuel removal wherever a control rod can be permanent-

ly or temporarily dispensed with. In such positions the sealing plugs of

the standpipes may be easily modified to enable insertion of experiment-

al equipment into the core. For instance, a pressure reentrant tube

can be joined to a plug through which a suitable access hole has been

drilled. This accessibility is a great advantage in connection with the

physics measurements.

3. SCOPE OF EXPERIMENTS

The possibility of making measurements on a full-sized system

with proper operational characteristics made way for a rather compre-

hensive reactor physics programme. The aim was to assess the physics

properties for a varies set of conditions with different core radii, mode-

rator levels, temperatures and power levels in order to check the theo-

retical models for the reactor physics calculations, in addition to the

determination of the required operational characteristics. We shall

briefly point out some of the most interesting results obtained in con-

nection with the startup experiments. Details of the experiments and

their results are given elsewhere f 4 - 7].

3.1 Buckling measurements

Buckling measurements imply the determination of the critical

moderator level (without control rods) and the corresponding thermal

neutron flux distribution as a function of temperature. It may be men-

tioned that because of pressurizing difficulties at low moderator levels,

high temperature buckling measurements on a control rod free core

have not yet been carried out.

Cold buckling results, as expected, agree well with those obtain-

ed from critical assembly zero power experiments. A higher excess

reactivity than expected was found, however. This is due, at least partly,

to the fact that the theory underestimated the effectiveness of the re-

flector. Since high temperature measurements could only be done with

a filled tank and therefore with banks of control rods inserted into the

core, the corresponding buckling value obtained is somewhat inaccurate.

- 6 -

The result, however, indicated agreement with information from high

temperature exponential experiments, verifying that the temperature

dependence of the buckling is less than calculated [4 ] . At full power,

excess reactivity was in fact about 1.5% higher than inferred from

theory. Such reactivity deviations are, of course, very important in

connection with natural uranium fuelled reactors [5 ] .

3. 2 Reactivity: .coefficients

The reactivity dependence on moderator level and temperature

as well as differential control rod worths was determined I" 61 by means

of period measurements employing the electronic data acquisition unit

RAMSES (to be described below). Special in-core high temperature

neutron counters were used since no external neutron source was in-

cluded and the source flux level (initially originating from spontaneous

fission only) was low before high power operation. Differential level

increments were attained with a high pressure piston pump with a vari-

able piston stroke (in the drainage-return auxiliary circuit). The pump

was calibrated against the special level meter. Temperature changes

were governed by regulating the power output of the secondary circuit

electric heater.

Changes in control rod positions were accurately determined by

merely counting the number of steps the rods move, each step being

10.0 jh 0. 1 mm. This is a result of the hydraulic drive unit characterist-

ics.

The moderator level coefficient measured for the clean (control

rod free) core in conjunction with critical level determinations for vari-

ous control rod configurations was used to calculate the control rod

worths. Because the equipment enabled us to make accurate measure-

ments, the experimental results allowed close comparisons with theo-

retical calculations. In particular, the three-dimensional heterogeneous

code, HETERO, recently developed in Sweden, has been shown to give

too low worths by some 5 to 15 %. The temperature coefficient measured

includes the control rod worth contribution, the effect of which is diffi-

cult to separate from the lattice coefficient in an accurate manner.

- 7 -

In spite of this, the results indicate the same excess reactivity tempe-

rature dependence as in the buckling measurements.

3.3 Control rod worths

Integral reactivity equivalences of control rods were assessed

in two ways. First, for a limited reactivity range, the reactivities may

be determined from the change in the critical level when rods are in-

serted into the core. This method pre-supposes a knowledge of the level

coefficient as a function of moderator level.

Second, subcritical measurements with the neutron pulse tech-

nique may be used to give shutdown reactivities for various conditions.

The method is favourable for large reactivity values and is time saving.

The comparison with calculations is facilitated by the fact that no geo-

metrical changes are induced by the measurement. In our case a draw-

back is the large size of the system and the necessity of making all

measurements before any notable core irradiation [7],

A practical consequence of the high effectiveness found for the

control rods was that the number of rods required for normal operation

could be reduced significantly in relation to the total number available.

3.4 Form factors

The control rods inserted during power operation distort the

flux distribution, giving generally unfavourable flux form factors. The

precise calculation of these factors is difficult, and sufficient experi-

mental information is lacking. In view of this, wire activation measure-

ments were carried out in a few cases with different control rod con-

figurations and temperature. Subsequent comparisons were made be-

tween the activation distributions and calculations based on hetero-

geneous methods.

Another means of assessing form factors is to record the rela-

tive temperatures of exit coolant from the channels. This has been done

by photographing the two oscilloscope screens on which all the tempera-

tures are displayed simultaneously. Fission product distributions in

fuel removed from the reactor -were also examined.

- 8 -

For the Agesta reactor a detailed knowledge of the form factors

is not essential for normal operation because of the relatively low

specific power. However, the measurements carried out are of interest

for comparison with the HETERO calculations mentioned above. Satis-

factory agreement has been found [4, 5]. <

3.5 Microscopic studies

Detailed neutron flux studies in a special fuel assembly were

carried out by means of foil activation techniques to determine the fuel

disadvantage factor, spacer gap flux peaking effects and spectral para-

meters [3] , The foils can be positioned either between the UO_ pellets,

and are then of the same diameter, 1.70 cm, as these, or in the coolant

between rods. Uranium metal foils were used to assess the fission rate

distributions. For the spectrum studies, foils with differing neutron

absorption cross-section dependence were activated. The activation

ratios between the various foil materials in the Agesta fuel in conjunc-

tion with measurements on known spectra (measured by time-of-flight

techniques) are hoped to furnish adequate information to yield a multi-

parameter description of flux spectra.

The experiments showed the spacer gap perturbation to be

slightly more pronounced, although of less range, than was anticipated

from calculations. As regards the spectral studies, the foil activation

measurements were of adequate precision to justify advanced compari-

sons with spectrum calculations, e.g. with the THERMOS code. This

work is still under way.O o Q

The U resonance to thermal activation ratio, a primary

quantity in the evaluation of the resonance escape probability and the

initial conversion ratio, has been measured by counting the Np-239

activity induced in UO~ pellets irradiated in the special fuel rods. The

l/v activation was monitored with Cu foils and the complementary

thermal column irradiation was done in the research reactor Rl f 8],

3.6 Dynamic studies

The assessment of transfer functions to determine kinetic para-

meters and stability characteristics involves the analysis of noise

„ 9 -

power spectra and reactivity oscillations. The Agesta reactor is parti-

cularly suited to test such methods since the reactivity feed-back

mechanisms are relatively simple, comprising mainly fuel and mode-

rator temperature coefficients.

The noise analysis employs both analogue and digital recording

techniques. In the former method the signal is recorded on magnetic

tape. Subsequent analysis makes use of a wave analyser, a voltage -

frequency converter and a sealer. For digital recording RAMSES is

used, giving all relevant data on punched tape ready for computer

analysis.

To induce reactivity perturbations, one of the regulating rods

is oscillated between adjustable limits. In addition to constant-period

trapezoid perturbations, two different types of binary multifrequency

programmes are applied. All quantities such as rod position, power,

temperatures and secondary steam parameters are recorded digitally

with the RAMSES for subsequent machine analysis.

So far the experience gained from the dynamics development

programme suggests that the combined use of multifrequency perturba-

tions, step functions, and power noise analyses provides an effective

method of determining the dynamic operational properties of the reactor.

The preliminary measurements have revealed some discrepancies be-

tween experiment and theory, especially with regard to the fuel power

coefficient and the fuel temperature time constant (which are, of course,

difficult to calculate accurately). In general, still better stability

characteristics as compared with analogy machine studies have been

found. A somewhat lower value of the effective heat transfer coefficient

in the primary heat exchangers than those calculated was also indicated

by the experiments Ï 5].

3. 7 Fuel burnup

The first fuel cycle will make use of a once-through batch and

will be carefully followed up with the aid of four special fuel assemblies.

A number of the fuel rods in these have been compared with a standard

rod in the zero-power R0-reactor by the pile oscillator method prior

to loading into the Âgesta core. The fuel assemblies are then irradiated

- 10 -

in different radial positions and some six months after removal from

the reactor the rods will once more be compared to the standard un-

irradiated rod. The maximum irradiation will be near 6000 MWd/ton.

This will yield the integral change in the production cross "Section of

the fuel, apart from spectral differences. The neutron dose is measured,

using niobium foils between the pellets. The irradiated fuel will also be

examined by mass and gamma-spectroscopic methods.

During the course of irradiation, neutron flux spectrum changes

in the coolant between fuel rods will be followed up by intermittent foil

activations. Such spectrum studies are believed to give important

information for cross-section calculations in connection with burnup

assessments.

4. EXPERIMENTAL EQUIPMENT

4.1 In-core detectors

With fresh fuel in a D_O-moderated reactor and relying only on

spontaneous fission and photoneutrons as a start-up source, the station-

ary external neutron fLux detectors have to be supplemented by a more

sensitive in-core system. This is especially true of cores having small

radii and large reflectors -where even an added external source would

not otherwise suffice. The required in-core detectors, which were used

until the photoneutron source was built up prior to the high power

approach, covered the power range up to a few kW whereby more than

adequate overlapping with the external detectors was obtained. The

neutron counters were required to operate at 220 C.

For housing the internal detectors, two identical stainless steel

thimbles with a design pressure of 40 bar at 220 C and a diameter of

44.5 mm were assembled in the core region (10 in fig. 2). Each thimble

contained one boron and one fission counter together with three re-

sistance thermometers, all mounted on a movable holder suspended in

a wire which was fastened at the top shielding cover (16 in fig. 2). The

radial thimble position could be chosen at will with due consideration

to control rod requirements.

- 11 -

The temperature transducers, which were in close thermal

contact with the inner thimble surface, were connected to a Wheat-

stone bridge by means of a three-lead connection so that lead resist-

ances could be measured separately. The thermometers were cali-

brated to within 0.1 C and have shown good reproducibility.

Detector response with boron chambers was noted with only

a single fuel assembly introduced into the core. This indicated the

natural neutron source to be adequate for approach to critical proce-

dures.

Numbering of components in fig. 2 (see next pagej

(numbers in brackets specify the number of units)

1 Normal fuel assembly (136)

2 Coarse control rod (18) at full insertion

3 Special fuel assembly (4) for burnup studies

4 Slot for insertion of foil holder between fuel rods

5 Special fuel assembly for fine structure measurements

6 Removable group of six fuel rods

7 Zircaloy wire holder (3) for axial flux distributionmeasurements

8 Activation wire (Cu or Co)

9 Activation wire attached to normal fuel assembly shroud (15)

10 High pressure thimble for neutron detectors

11 High pressure thimble for neutron pulse generator

12 D2O level gauge

13 External tube (3) for flux measurements at tank surface

14 Preamplifiers for neutron detectors

15 Standpipe for fuel charging and housing of control rodmechanisms

16 Iron shielding cover

17 Shielding plug

18 Top concrete shielding (used at high power)

Fig. 2 Cross sectional view of theAgesta power reactor showingin-core instrumentation forphysics experiments.

- 13 -

4 . 2 level meter

For the start-up experiments an accurate level meter was

needed for the determination of critical levels within a few milli-

metres. Since the stationary equipment was far too uncertain, a special

equipment was designed. Working with a steady nitrogen gas flow, it

measures the differential pressure between two pipes, their orifices

being above and below the water level respectively (cf. fig. 3). The

submerged orifice, expelling bubbles into the moderator, then ex-

periences a pressure difference equivalent to the displaced water

column. In practice the meter, which is assembled from conventional

SS components, has six tubes entering the reactor tank through the

sealing plug of a control rod free standpipe (12 in fig. 2). This arrange-

ment yields five level ranges of 50 cm each, whereby the quoted dp-cell

accuracy of +_ 0. 5 % corresponds to +_ 0. 25 cm in water level.

dp-transmitter

Recorder

Diff. press regulator

Fig. 3 Principle of moderator levelgauge performance. The range oflevels covered is from 120 to370 cm (filled tank) at all pressures.

Referring to fig. 3, the N -gas pressure is applied to two near-

identical piping branches, the pressure being 3 to 5 bar above the

- 14 -

reactor vessel value. In each branch there are regulating valves to

ensure equal gas flows. Further, there are differential-pre s su re regu-

lators that automatically maintain a steady gas flow irrespective of the

pressure and the level changes in the reactor. Under these conditions

the recorded pressure change will depend only on moderator level

changes, provided that moderator streaming effects are small.

\

\\

\

"A ' % •

Fig. 4 Equipment located on top shielding. The level gaugetransmitter and associated components are shownmounted on the stand. The level ranges are selectedby means of the magnetic valves operated from thebox on the left. At right the preamplifiers for theneutron detectors.

A pneumatically operated dp-transmitter (Taylor Instruments Co)was chosen (see fig. 4) owing to its higher accuracy (+0.5 %) as com-pared with corresponding electrical systems (at best +_ 1 %). All valvesexcept those for regulating are electromagnetically controlled. This re-sults in convenient centralized operation of the equipment. A means of

- 15 -

calibrating the transmitter is to read the pressures for two adjacent

submerged orifices, when possible. The pressure difference then

corresponds to an accurately known -water column. A short-circuit

connection permits zeroing of the transmitter. Consumption of N_-gas

amounts to some 30 lit. /h at atmospheric pressure and rises in pro-

portion to pressure.

The overall absolute accuracy is estimated to range from

Jh 0.2 to +_ 0.4 cm with rising pressure. Leaving out systematic errors,

the uncertainty will be about 0. 1 cm less.

From our experience we conclude that the level meter, although

somewhat robust in design, is fully suited to yield accurate critical

level determinations. With our system tightness and leakage problems

associated with a movable shaft entering a pressurized system, as for

instance with dipsticks, are wholly avoided.

4.3 Pulsed neutron source

A pulsed neutron source, Kaman type NT 60-7, was used for

the measurement of reactivities at various subcritical states. The

neutron yield was about 10 n/pulse with this specific source.

Two thimbles for the neutron source were used for low and high

temperatures respectively. The reason for this was a maximum source

operating temperature of 70 C and the relative bulkiness of the gene-

rator (O.D. 4" x 20"). With the low temperature/low pressure thimble

the amount of structural material inserted into the core could be kept

at a minimum (tube wall thickness 0. 3 cm). In this thimble the neutron

source could be lowered to 120 cm above the bottom plate by means of

a suspension wire.

The high temperature thimble (11 in fig. 2) to be used up to

220 C was provided with a light water cooler, see fig. 5. The maximum

source insertion depth (250 cm above bottom plate) in the core was a

compromise between source effectiveness and amount of material in-

troduced.

As is shown in fig. 5, the cooler is made of two concentric

tubes axially partitioned into two halves. The cooling water flows down-

wards in one half and upwards in the other. At a cooling water circula-

- 16 -

tion rate of 3 lit./min., maximum air temperature inside the cooler was

27 C at 220 C moderator temperature. Coolant failure was arranged to

give an alarm in the control room. The coolant container was long

enough to allow a choice of source positions inside it over a distance

of 50 cm by means of a suspension wire.

The neutron pulses were triggered from a control unit (placed

on the top shielding), which in turn was operated from the control room

by trigger pulses from a 256-channel TMC time-analyzer and a control

unit. The accumulated data was printed out as well as punched on tape

for subsequent computer analysis.

4.4 Activation wire holders

The neutron flux measurements employ -wires mounted vertically

on the outside of ordinary fuel assembly sheaths (9 in fig. 2) or in

special holders placed in control rod positions (7 in fig. 2), i .e. in the

moderator, equidistant from four adjacent fuel assemblies.

Fifteen fuel assemblies have thus been fitted with wire supports

at the top and bottom ends and are used with 0. 1 cm diam. Cu wires

(electrolytic quality) for low power measurements. A practical power

limit (âî 10 kW) is set by radiation handling considerations.

The three special holders in "moderator positions" are also

intended for use at.full power using wires of 0. 5 % Co alloyed with Al.

These holders are handled with the charging machine. The activation

wire is attached and released remotely with a special tool when the

wire holder is placed in a control rod drive storage position. In this

way no special radiation protection measures have to be taken.

The wire holder is formed in two parts. The bottom part, ex-

tending through the core, is a Zircaloy tube similar to the fuel assembly

sheaths. The bottom end of this tube is fitted with a guide cone for posi-

tioning in the bottom grid plate (see fig. 2) to avoid swinging movements

induced by moderator streaming. In the cone is a wire support of

bayonet catch type with holes to ensure proper filling or emptying of

water.

When the holder is inserted into the reactor, even a small mis-

alignment would cause the bottom cone to miss the circle of 4.6 cm

- 17 -

Bottom end of:or standpipe

, AtmosphericI pressure

coolant flow tubes3lit/min exit temp.<30*C

Fig. 5 Arrangement for housing andcooling of neutron pulsegenerator in the core.

- 18 -

diam. into which it has to be guided. For this reason the top part of

the holder, located in the standpipe, is coupled to the bottom part by

means of a gyro suspension. This suspension also allows a certain rela-

tive axial movement between the two parts to account for differential

expansion with temperature and pressure changes. With this design

the holder tube at all times stands on the bottom grid plate by its own

weight.

The wire, which is stretched along the tube axis, is connected

at the top to a support constituting one half of a bayonet joint and pro-

vided with a spring to take up differential expansion between the wire

and the tube.

The tool for attaching and detaching the wire is very simple,

consisting mainly of an iron tube (02.2 x 0. 175 x 520 cm) containing

an iron rod, the bottom of which is fitted with a bayonet catch. Briefly,

the procedure for wire removal is first to guide the central rod onto

the top wire support and catch hold of it; then insert the tube, at the

same time enclosing the top support and the wire itself until the bottom

support is reached and is released with the tube. The tool, now con-

taining the activation wire, can then be removed. A reverse procedure

is used for loading the holder with a new wire.

On removal, the wires are gamma-counted on an automatic

wire scanner with a scintillation counter to which the electronic timing

and counting register of RAMSES is connected. The output is obtained

in the form of punched tape for subsequent computer analysis.

4. 5 Special fuel assembly for fine structure studies

For the measurements of neutron flux spectra and fine structure

in the fuel and ordinary fuel assembly was modified to include a remov-

able group of six representative fuel rods in one of the four 19-rod

clusters (5 and 6 in fig. 2 and fig. 6). The rods are normally screwed

into each other axially, but in the modified section (second from bottom)

the six experimental rods can be detached laterally through a rectangu-

lar opening in the coolant tube. The curved plate cut out from the tube

is fitted with a top and bottom grid, thus forming a rod holder. When

inserted, the holder is secured with two lock bolts. In spite of the

- 19 -

relatively large mechanical changes, the assembly has the same normal

functional data and amount of structural material (Zircaloy).

The fuel assembly sheath was also provided with eight slots,

0.25 x 7. 1 cm, two along each of the four sections. By this means, foil

holders in the form of sheets can be inserted between the fuel rods

along two "diameters" in the manner illustrated in fig. 6 c. The holders

are actually composed of two separate Zircaloy sheets (11.0 x 7.0 x

0.06 cm), one of them having five milled tracks for foil positioning.

The fuel assembly would be irradiated typically at 10 n/cm s

(some 10-20 kW of power) for 1.5 hours. Since the active substance in

some of the foils is very thin to avoid resonance self-shielding, it is

desirable to remove the foils from the fuel as soon as possible for

counting before decay becomes appreciable. On the other hand, there

is the necessity to wait for the discharge of the assembly from the

reactor to allow for cooling down after a high temperature irradiation

or, after low temperature irradiation, for the gamma ray dose rate

from the fuel to diminish sufficiently.

For the removal of the experimental pins from the fuel assembly,

this is lowered from the charging machine into the space behind a special

iron shielding (10 cm thick).

Into this shield has been built a manipulator for detaching the

rod holder from the assembly. Such an arrangement is necessary since

the surface dose rate may amount to 30 r/h. The sheet foil holders are

removed with a separate manipulator, as will be described below.

Subsequently the fuel rods, which have been welded tight prior

to an irradiation, must be opened up. For this procedure each rod is

in turn encased in a 20 cm diameter cylindrical iron shield with a 1.9 cm

diameter hole along the axis. The top plug of the fuel rod is sawn off

and an axial hole drilled through the bottom plug, enabling the pellets

and foils to be pushed out.

4.6 Equipment for burnup studies

The four assemblies (item 3 in fig. 2) to be used for burnup

experiments include 46 experimental rods so distributed between them

as to yield a suitable range of burnup values. The niobium foils used

- 20 -

Fig. 6a Details of special fuel assemblyfor fine structure studies:top section of ordinary assembly.

- 21 -

6b Details of special fuel assembly forfine structure studies:removable set of six fuel rods froman intermediate section.

- 22 -

LûckJiûiL

Experimentalfuel pins (6)

Fig. 6c Details of specialfuel assembly forfine structurestudies :location of experimentalrods and foil holder.

for monitoring the integral neutron dose are placed between some of

the UO_ pellets, three in each position, the outer two serving as fission

product shields.

Each assembly also has two slots in each of the two top sections.

Here, foil holders similar to those used in the "fine structure assembly"

can be inserted. The number of slots was limited to four in each

assembly to avoid excess coolant leakage to the moderator.

For the insertion and removal of foil holders from the highly

irradiated fuel assemblies, a special manipulator is used (see fig. 7).

The manipulator has been used at a previous service station for ex-

change of assembly orifices, allowing the fuel to be handled by the

charging machine only. Fairly simple equipment can then be used.

Besides providing effective radiation shielding, fuel cooling is

also attended to without any special measures. Referring to fig. 7,

- 23 -

Mechanism far locking the6»Lassfimbly _ _ , Tongs foe, -

foil holds) '/ \ •/////

ManjfuiLalflt Jot-insertionbf Jpil holder m to the ]fuel cluster __

Fig. 7 Device for positioning foil holdersin irradiated fuel assemblies forburnup studies.

the manipulator comprises in principle a pair of pliers for gripping

the foil holder and a clutch which fastens the assembly when the coolant

tube slot is aligned with the pliers. The clutch and the pliers are oper- •

ated by shafts penetrating the shielding. Built-in lamps provide light

for viewing the operations through a periscope.

An axial gamma ray scan of the fuel for assessing the burnup

distribution along a fuel assembly can easily be performed in the same

service station by inserting a collimator in place of the manipulator,

and placing a scintillator detector in front of it.

4.7 Reactivity oscillator

Either or both of the two regulating rods are used to induce

reactivity oscillations. The rods have hydraulic drives of piston type

and can be operated sufficiently fast for oscillating purposes. Their

differential reactivity worth is also adequate to yield optimum oscilla-

ting conditions. At the same time the total rod worth is limited to fulfil

all safety demands.

- 24 -

The oscillations are programmed on a binary basis, i .e. an

arbitrary but definite amplitude is used in a specific measurement

whether of the trapezoid or multifrequency type. A time relay deter-

mines the stroke of the regulating rod from full insertion to the parti-

ally withdrawn position. The programme unit is of tape-reader type

and supplies an arbitrary binary perturbation signal which acts on the

regulating rod through a relay.

Typically, the regulating rod stroke is 5-20 cm corresponding

to a reactivity amplitude of 25 - 90 • 10" . The minimum unit time

interval chosen (2 sec.) is limited by the duration of a stroke; maxi-

mum rod speed is 15 cm/s.

4.8 The data acquisition unit RAMSES

As should be evident from the foregoing the RAMSES unit has

proved extremely valuable for our physics measurements. It is a flex-

ible all-transistorised system for automatic collection and recording of

of digital or analog data [[43 -

Fig. 8 The data acquisition unitRAMSES.

- 25 -

Thus the input signals may originate in, e.g. pulse detectors

or in analogue-to-digital converters. The numbers counted are trans-

ferred to an output device, a tape reperforator, a typewriter or

a magnetic tape recorder.

The instrument comprises three main units, a control unit for

reading and printing the information, a programme and a register unit.

The versatility of the equipment shows up in the plug-in board modular

system for these latter units. Both the programme and the register

units contain 32 positions in which standardised circuits on plug-in

boards may be inserted. In the register unit the required number of

sealers are set up together with instruction character boards. In the

programme unit the desired mode of automatic operation is determined

by a timer and the logic circuit board combination, in addition to which

one may incorporate linear pulse amplifiers, pulse height discrimi-

nators, an analogue-to-digital converter and a signal selector for

simultaneous multiple data assessment of analogue signals. The con-

trol unit, finally, transfers the sealer information to some output

device. A clock built into the unit supplies pulses to the timer.

A simple application of the RAMSES is its utilization for period

measurements or for recording reactor power generally. The input

from each of two pulse counters (the in-core detectors, for instance)

is counted alternately in two six-digit sealers at preset time intervals

and the output is recorded on tape. This method avoids the time lag

for readings between successive time intervals.

A rather more complicated application is the data recording

in connection with the dynamics measurements, where the capacity of

the RAMSES is more fully exploited. During the reactivity oscillations,

up to eight analogue signals can be sampled intermittently by means of

a signal selector which records the simultaneous signals in a condenser

memory. The signals are then read off successively and are transfer-

red to a sealer through an analogue-to-digital converter. In addition

to these signals, two digital signals can be recorded as -well as the

ideal perturbation signal.

- 26 -

1. McHUGH B (ed.)»The Âgesta Nuclear Power Station;AB Atomenergi, Stockholm 1964

2. RYDELL N, BLOMBERG P E, ERICSSON E,Experience from the commissioning, the criticality experimentsand the power operation of the Agesta Nuclear Power Plant.Int. Conf. Peaceful Uses Atomic Energy. 3, 1964, Geneva.Proc. Vol. 5, 1964, p. 336

3. • SANDSTRÖM S,Operating experience at the Âgesta Nuclear Power Station. 1966.( 6 )

4. APELQVIST G, BLOMBERG P E, e t a l . ,Reactor physics studies and comparisons between reactor physicsdata from calculations and mock-up studies and from measure-ments in the Âgesta Nuclear Power Plant.Int. Conf. Peaceful Uses Atomic Energy. 3, 1964, Geneva.Proc. Vol. 3, 1964, p. 458

5. APELQVIST G, BLISELIUS P-Â, BLOMBERG P E, e t a l . ,Physics experiments at the Âgesta Power Station. 1.966.(AE-2'44)

6. BERNANDER G,AE-report to be published.

7. BJÖREUS K,AE-report to be published.

8. BERNANDER G,AE-report to be published.

LIST OF PUBLISHED AE-REPORTS

1-190. (See the back cover earlier reports.)191. Trace elements in the conductive tissue of beef heart determined by

neutron activation analysis. By P. O. Wester. 1935. 19 p. Sw. cr. 8 : - .192. Radiolysis of aqueous benzene solutions in the presence of inorganic

oxides. By H. Christensen. 12 p. 1965. Sw. cr. 8 : - .193. Radiolysis of aqueous benzene solutions at higher temperatures. By H.

Christensen. 1965. 14 p. Sw. cr. 8 : - .194. Theoretical work for the fast zero-power reactor FR-0. By H. Haggblom.

1965. 46 p. Sw. cr. 8 : - .195. Experimental studies on assemblies 1 and 2 of the fast reactor FRO.

Part 1 . By T. L. Andersson, E. Hellstrand, S-O. Londen and L. I. Tiren.1965. 45 p. Sw. cr. 8 : - .

196. Measured and predicted variations in fast neutron spectrum when pene-trating laminated Fe-D:O. By E. Aalto, R. Sandlin and R. FrSki. 1965. 20 p.Sw. cr. 8 : - .

197. Measured and predicted variations in fast neutron spectrum in massiveshields of water and concrete. By E. Aal io, R. Fraki and R. Sandlin. 1965.27 p. Sw. cr. 8 : - .

198. Measured and predicted neutron fluxes in , and leakage through, a con-figuration of perforated Fe plates in D;O. By E. Aalto. 1965. 23 p. Sw.cr. 8 : - .

199. Mixed convection heat transfer on the outside of a vertical cylinder. ByA. Bhattacharyya. 1965. 42 p. Sw. cr. 8 : - .

200. An experimental study of natural circulation in a loop with parallel f lowtest sections. By R. P. Mathisen and O. Eklind. 1965. 47 p. Sw. cr. 8 : - .

201. Heat transfer analogies. By A. Bhattacharyya. 1965. 55 p. Sw. cr. 8 : - .202. A study of the "384" KeV complex gamma emission from plutonium-239.

By R. S. Forsyth and N. Ronqvist. 1965. 14 p. Sw. cr. 8 : - .203. A scinti l lometer assembly for geological survey. By E. Dissing and O.

LandstrBm. 1965. 16 p. Sw. cr. 8 : - .204. Neutron-activation analysis of natural water applied to hydrogeology. By

O. Landstrom and C. G. Wenner. 1965. 28 p. Sw. cr. 8 : - .205. Systematics of absolute gamma ray transition probabil i t ies in deformed

odd-A nuclei. By S. G. Malmskog. 1965. 60 p. Sw. cr. 8 : - .206. Radiation induced removal of stacking faults in quenched aluminium. By

U. Bergenlid. 1965. 11 p. Sw. cr. 8 : - .207. Experimental studies on assemblies 1 and 2 of the fast reactor FRO. Part 2.

By E. Hellstrand, T. Andersson, B. Brunfelter, J . Kockum, S-O. Londenand L. I. Tiren. 1965. 50 p. Sw. cr. 8 : - .

208. Measurement of the neutron slowing-down time distribution at 1.46 eVand its space dependence in water. By E. Moller. 1965. 29 p. Sw. cr. 8 : - .

209. Incompressible steady f low with tensor conductivity leaving a transversemagnetic f ie ld . By E. A. Wital is. 1965. 17 p. Sw. cr. 8 : - .

210. Methods for the determination of currents and fields in steady two-dimensional MHD flow with tensor conductivity. By E. A. Wital is. 1965.13 p. Sw. cr. 8 : - .

211. Report on the personnel dosimetry at AB Atomenergi during 1964. ByK. A. Edvardsson. 1966. 15 p. Sw. cr. 8 : - .

212. Central reactivity measurements on assemblies 1 and 3 of the fast reactorFRO. By S-O. Londen. 1966. 58 p. Sw. cr. 8 : - .

213. Low temperature irradiation applied to neutron activation analysis ofmercury in human whole blood. By D. Brune. 1966. 7 p. Sw. cr. 8:—.

214. Characteristics of linear MHD generators with one or a tew loads. ByE. A. Wital is. 1966. 16 p. Sw. cr. 8 : - .

215. An automated anion-exchange method for the selective sorption of fivegroups of trace elements in neutron-irradiated biological material. ByK. Samsahl. 1966. 14 p. Sw. cr. 8 : - .

216. Measurement of the time dependence of neutron slowing-down and therma-lization in heavy water. By E. Moller. 1966. 34 p. Sw. cr. 8 : - .

217. Electrodeposition of actinide and lanthanide elements. By N-E. Barring.1966. 21 p. Sw. cr. 8 : - .

218. Measurement of the electrical conductivity of He3 plasma induced byneutron irradiation. By J. Braun and K. Nygaard. 1966. 37 p. Sw. cr. 8:—.

219. Phytoplankton from Lake Magelungen, Central Sweden 1960-1963. By T.Wi l len. 1966. 44 p. Sw. cr. 8 : - .

220. Measured and predicted neutron flux distributions in a material surround-ing av cylindrical duct. By J. Nilsson and R. Sandlin. 1966. 37 p. Sw.cr. 8 : - .

221. Swedish work on brittle-fracture problems in nuclear reactor pressurevessels. By M. Grounes. 1966 34 p. Sw. cr. 8:—.

222. Total cross-sections of U, UOi and TI1O2 for thermal and subthermalneutrons. By S. F. Beshai. 1966. 14 p. Sw. cr. 8 : - .

223. Neutron scattering in hydrogenous moderators, studied by the time de-pendent reaction rate method. By L. G. Larsson, E, Moller and S. N.Purohit. 1966. 26 p. Sw. cr. 8 : - .

224. Calcium and strontium in Swedish waters and f ish, and accumulation ofstrontium-90. By P-O. Agnedal. 1966. 34 p. Sw. cr. 8 : - .

225. The radioactive waste management at Studsvik. By R. Hedlund and A.Lindskog. 1968. 14 p. Sw. cr. 8 : - .

226. Theoretical t ime dependent thermal neutron spectra and reaction ratesin HiO and D2O. S. N. Purohit. 1966. 62 p. Sw. cr. 8 : - .

227. Integral transport theory in one-dimensional geometries. By I. Carlvik.1966. 65 p. Sw. cr. 8 : - .

228. Integral parameters of the generalized frequency spectra of moderators.By S. N. Purohit. 1966. 27 p. Sw. cr. 8 : - .

229. Reaction rate distributions and ratios in FRO assemblies 1, 2 and 3. ByT. L. Andersson. 1966. 50 p. Sw. cr. 8 : - .

230. Different activation techniques for the study of epithermal spectra, app-l ied to heavy water lattices of varying fuel-to-moderator ratio. By E. K.Sokolowski. 1966. 34 p. Sw. cr. 8 : - .

231. Calibration of the failed-fuel-element detection systems in the Agestareactor. By O. Strindehag. 1966. 52 p. Sw. cr. 8 : - .

232. Progress report 1965. Nuclear chemistry. Ed. by G. Carleson. 1966. 26 p.Sw. cr. 8 : - .

233. A Summary Report on Assembly 3 of FRO. By T. L. Andersson, B. Brun-felter, P. F. Cecchi, E. Hellstrand, J. Kockum, S-O. Londen and L. I.Tiren. 1966. 34 p. Sw. cr. 8 : - .

234. Recipient capacity of Tvaren, a Baltic Bay. By P.-O. Agnedal and S. O. \NBergstro'm. 21 p. Sw. cr. 8 : - .

235. Optimal linear fi lters for pulse height measurements in the presence olnoise. By K. Nygaard. 16 p. Sw. cr. 8 : - .

236. DETEC, a subprogram for simulation of the fast-neutron detection process in a hydro-carbonous plastic scinti l lator. By B. Gustafsson and OAspelund. 1963. 26 p. Sw. cr. 3 : - .

237. Microanalys of fluorine contamination and its depth distribution in zircaloyby the use of a charged particle nuclear reaction. By E. Motler and NStarfelt. 1966. 15 p. Sw. cr. 8 : - .

238. Void measurements in the regions of sub-cooled and low-quality boi l ingP. 1. By S. Z. Rouhani. 1966. 47 p. Sw. cr. 8 : - .

239. Void measurements in the regions of sub-cooled and low-quality boi l ingP. 2. By S. Z. Rouhani. 1966. 60 p. Sw. cr 8 : - .

240. Possible odd parity in m X e . By L Broman and S. G. Malmskog 193510 p. Sw. cr. 8 : - .

241. Burn-up determination by high resolution gamma spectrometry spectrafrom slightly-irradiated uranium and plutonium between 400-830 keV ByR. S. Forsyth and N. Ronqvist. 1966. 22 p. Sw. cr. 8 : -

242. Half l i fe measurements in " s G d . By S. G. Malmskog. 1966. 10 p. Swcr. 8 : - .

243. On shear stress distributions for flow in smooth or partially rough annuliBy B. Kjellstrom and S. Hedberg. 1966. 66 p. Sw. cr. 8 . - .

244. Physics experiments at the Agesta power station. By G. Apelqvist, P -ABlisel ius, P. E. Blomberg, E. Jonsson and F. Akerhielm. 1966. 30 p Swcr. 8 : - .

245. Intercrystalline stress corrosion cracking of inconel 600 inspection tubes inthe Agesta reactor. By B. Gronwall, L. Ljungberg, W. Hubner and WStuart. 1966. 26 p. Sw. cr. 8 : - .

246. Operating experience at the Agesta nuclear power station. By S Sand-strom. 1966. 113 p. Sw. cr. 8 : - .

247. Neutron-activation analysis of biological material with high radiation levelsBy K. Samsahl. 1968. 15 p. Sw. cr. 8 : - .

248. One-group perturbation theory applied to measurements with void. By RPersson. 1966. 19 p. Sw. cr. 8 : - .

249. Optimal linear f i l ters. 2. Pulse time measurements in the presence olnoise. By K. Nygaard. 1966 9 p. Sw. cr. 8 : - .

250. The interaction between control rods as estimated by second-order one-group perturbation theory. By R. Persson. 1966 42 p. Sw. cr. 8*-

251. Absolute transition probabil i t ies from the 453.1 keV level in 183W. By S GMalmskog. 1966. 12 p. Sw. cr. 8 : - .

252. Nomogram for determining shield thickness for point and line sources olgamma rays. By C. JBnemalm and K. Malen. 196S. 33 p. Sw. cr. 8 . -

253. Report on the personnel dosimetry at AB Atomenergi during 1985 By K AEdwardsson. 19S6. 13 p. Sw. cr. 8 : - .

254 Buckling measurements up to 250°C on lattices of Agesta clusters and onD2O alone in the pressurized exponential assembly TZ. By R. Persson,A. J. W. Andersson and C.-E. Wikdahl. 1966. 56 p. Sw. cr. 8 : - .

255 Decontamination experiments on intact pig skin contaminated with beta-gamma-emitting nuclides. By K. A. Edwardsson, S. Hagsgard and A Swens-son. 1966. 35 p. Sw. cr. 8 : - .

256. Perturbation method of analysis applied to substitution measurements ofbuckling. By R. Persson. 1966. 57 p. Sw. cr. 8 : - .

257. The Dancoff correction in square and hexagonal lattices. By I. Carlvik. 196835 p. Sw. cr. 8 : - .

253. Hall effect influence on a highly conducting f lu id. By E. A. Wital is. 196813 p. Sw. cr. 8 : - .

259. Analysis of the quasi-elastic scattering of neutrons in hydrogenous liquids.By S. N. Purohit. 1966. 26 p. Sw. cr. 8 : - .

230 High temperature tensile properties of unirradiated and neutron irradiated20Cr-35Ni austenitic steel By R B Roy and B Solly. 1986. 25 p. Swcr. 8 : - .

261. On the attenuation of neutrons and photons in a duct f i l led with a helicalplug. By E. Aalto and A. Krel l . 1986. 24 p. Sw. cr. 8 : - .

262. Design and analysis of the power control system of the fast zero energyreactor FR-O. By N. J. H. Schuch. 1966. 70 p. Sw. cr. 8 : - .

263. Possible deformed states in " s l n and " ' I n . By A. Backlin, B. Fogelberg andS. G. Malmskog. 1967. 39 p. Sw. cr. 10 : - .

264. Decay of the 16.3 min. " ! T a isomer. By M. Hojeberg and S. G. Malmskog.1967. 13 p. Sw. cr. 10: - .

265. Decay properties of " 7 N d . By A. Backlin and S. G. Malmskog. 1987. 15 p.Sw. cr. 10: - .

266. The half l i fe of the 53 keV level in " ' P t . By S. G. Malmskog. 1967. 10 p.Sw. cr. 10:—.

267. Burn-up extermination by hight resolution gamma spectrometry: Axial anddiametral scanning experiments. By R. S. Forsyth, W. H. Blackadder andN. Ronqvist. 1987. 18 p. Sw. cr. 10 : - .

268. On the properties of the s , / 2 >• d3y2 transition in " *Au . By A. Backlinand S. G. Malmskog. 1967. 23 p. Sw. cr. 10:- .

2S9. Experimental equipment for physics studies in the Agesta reactor. By G.Bernander, P. E. Blomberg and P.-O. Dubois. 1967. 35 p. Sw. cr. 10: - .

Forteckning over publicerade AES-rapporter

1. Analys medelst gamma-spektrometri. Av D. Brune. 1961. 10 s. Kr 6 : - .2. Bestralningsforandringar och neutronatmosfar i reaktortrycktankar - nagra

synpunkter. Av M. Grounes. 1962. 33 s. Kr 6 : - .3. Studium av strackgransen i miukt stal . Av G. Ostberg och R. Attermo

1963. 17 s. Kr 6 : - .

4. Teknisk upphandling inom reaktoromradet. Av Erik Jonson. 1963. 64 s.Kr 8 : - .

5. Agesta Kraftvarmeverk. Sammanstallning av tekniska data, beskrivningarm. m. for reaktordelen. Av B. Lill iehSok. 1964. 336 s. Kr 15:- .

6. Atomdagen 1965. Sammanstallning av foredrag och diskussioner. Av SSandstrom. 1966. 321 s. Kr 15:- .

Addit ional copies available at the library of AB Atomenergi, Studsvik, Ny-koping, Sweden. Micronegatives of the reports are obtainable through Film-produkter, Gamla landsvagen 4, Ektorp, Sweden.

EOS-tryokerierna, Stockholm 1967