5ATING CHARACTERISTICS CEN/SCK - IAEA

rSTUDIECENTRUM VOOR KERNENERGIE

CENTRE

TUD

DE

ENERG

NUCLEAIRE 5ATING CHARACTERISTICS

W. HEBELJ. PLANQUART

G. VANMASSENHOVE

CEN/SCKTECHNOLOGY AND ENERGY

DEPARTMENTMARCH 1976

BOERETANG 200

2400 MOL

BELGIUM

C.E.N./S.C.K.

TECHNOLOGY AND ENERGYDEPARTMENT

NT. 81/0180/64/WH/GVM/JP

IRRADIATION DEVICES FOR BR2 REACTOR

DESIGN AND OPERATING CHARACTERISTICS

by

W. HEBEL *

J. PLANQUART "

G. VANMASSENHOVE **

EURATOM

C.E.N7S.C.K.

March 1976

This report has been established in frame of the contract 010-734 BR2 Bbetween EURATOM and CEN/SCK on the utilization of BR2 reactor.

Contents

1 . General

2. Devices for Light Water Reactor (LWR) Development

2.1. Open Baskets (P-type)2.2. CEB2.3. PWC2.4. VNS

3. Devices for Liquid Metal cooled Fast BreederReactor (LMFBR) Development

3.1.3.2.3.3.3.4.3.5.3.6.3.7.

DISPCIRCECFCPOMFAFNIRFASOLDIPSL Loops

4. Devices for Gas cooled Breeder Reactor (GBR)Development

4.1. GSB Loop

5. Devices for High Temperature gas cooled Reactor(HTR) Development

5.1. CPR

6. Devices for Isotope Production

6.1. Open Baskets (R-type)6.2. Isotope, Capsules (CSF and IR)6.3. Cobalt Irradiation Devices6.4. Thimble Tubes (DG 36 and DGR)6.5. Hydraulic Rabbit (HR)6.6. Pool Side Tubes

19212529333739

45

45

47

49

53

555761636567

7. Operating and Irradiation Characteristics of BR2 69

4.

1. GENERAL

Since 1963 the material testing reactor BR2 at MOL(Belgium) is operated for the realisation of numerousresearch programs and experiments on the behaviour ofmaterials under nuclear radiation and in particularunder intensive neutron exposure.

During this period/special irradiation techniques andexperimental devices were developed according to thedesiderata of the different experiments and to theirradiation possibilities offered at BR2. The designand the operating characteristics of quite a number ofthose irradiation rigs of proven reliability may be usedor can be made available for new irradiation experiments.

A survey is given on the irradiation devices designedand constructed by CEN/SCK, Technology and Energy Dpt.and being currently used in BR2 at present. They arecompiled according to their main use for the differentresearch and development programs realised at BR2. Theirapplication however for different objectives can beenvisaged. No mention is given of irradiation devicesthat have been used only sporadically in the past orfor special terminated projects.

A final chapter summarizes the principal operating andirradiation conditions offered by BR2 reactor.

5.

2. DEVICES FOR LIGHT WATER REACTOR (LWR) DEVELOPMENT

Different irradiation rigs are available for testing thebehaviour of fuel rods being developed for use in water cooledreactors. The design and operating features of the inpiletest sections vary from simple non-instrumented baskets toelaborated rigs allowing to reproduce typical operatingconditions for the specimens as concern rod power, claddingtemperature, coolant pressure, power cycling, burn-up etc.

The rigs are in general of "capsule" type, meaning that thenuclear heat produced by the samples is dissipated to the BR2cooling water in a stationary manner, by radial heat transmission(conduction and convection) through the capsule wall structure,instead of dynamic cooling by forced convection as in a "loop"type device. Only in case of the so-called "Open Baskets" thefuel rods are directly cooled by the BR2 cooling water.The irradiations for LWR development are realised in reflectorchannels of BR2 using the flux of thermal neutrons being mainlyimportant for those tests.

I//7*867

7V761

•©

7-650 M " /

LONGITUDINAL SECTION 1111 I j

. ©SPECIAL AL FILLING PLUG

SS NEUTRON SCREEN

COOLING WATER (\\

BASKET TUBE /

LIFTINGEYE

CHANNELSHUT-OfF

PLUG

SUSPENSETUBE

7. 867 .

IRRADIATIONBASKET

USUAL Bê^ FILLING PLUG

J,

ICROSS SECTION AT CORE MID PLANE IVtO)

OPEN BASKET ( P-type)

7.

2.1. OPEN BASKETS (P-type)

Principal uses :

Testing of prototype fuel rods for PWR*and BWR*applicationat elevated power ratings and burn-up values;Reirradiation and up-rating of fuel rods with a certainburn-up coming from power reactors.

Design features :

Cf. Figure 2.1.

The P-type open basket devices are characterized by simplicityand were especially developed for reducing the preparatorydelay of an experiment to a minimum.The fuel rod is centered in an aluminium tube basket and isdirectly cooled by the BR2 primary cooling water.Neutron dose activation monitors and selfpowered neutrondetectors may be assembled with the basket to follow irra-diation history.Water sampling at the exit of irradiation channel can beperformed for cladding rupture detection.An automatic axial displacement mechanism may be assembledto the rig head for power cycling of the irradiation specimen.The reloadable basket allows intermediate hot cell inspectionsof the fuel rods during the irradiation campaign.By means of absorber tubes which can be placed around theirradiation basket, it is possible to reduce the thermalneutron flux in order to limit the rod power for elevatedfuel enrichments.

Operating characteristics :

Maximum fuel rod powerUsual fuel enrichmentFuel rod diameterActive fuel rod length, max.Coolant temperatureCoolant pressure, max.Irradiation in a reflector channel,typical unperturbed thermal neutronfluxOutside diameter of test section

700nat.9 . .70040 .14

(900). . 5%. 15

. . 50

. 1.5.

W/cm

mmmm°Cbar

14 910 n/cm^s

29,5 mm

Outpile control equipment :

According to the experimental desiderata specified, it can bemade available : a water sampling station, recorders for salf-powered neutron detectors, control panel for axial displacementor periodic cycling.

*PWR - Pressurized Water Reactor*BWR - Boiling Water Reactor

•VENT

IN P ILE SECTION

CROSS SECTIONAT CORE MIO PLANE

Fig2.2 CEB ( Capsule à eau bouillante)

9.

2.2. CEB

"Capsule à eau bouillante"

Principal uses :

Testing of prototype fuel rods for PWR application at elevated,typical cladding temperatures.

Design features :

Cf. Figure 2.2.

The fuel rod is surrounded by eutectic Sodium-Potassium alloy(NaK) contained in a copper capsule of cruciform cross section.A boiling stagnant water annulus at the outside of this capsuleallows a certain temperature regulation of- the fuel claddingby varying the boiling pressure of the water annulus. By meansof thermocouples attached to the copper capsule the claddingtemperature and the fuel rod heating rate can be determined.For special cases, cladding rupture of the fuel rod may bemonitored by measuring the pressure evolution in the gasplenum of the copper capsule.The rig can be charged into a He3 screen for power cycling of thefuel rods (cf. "VNS").


Maximum fuel rod powerUsual fuel enrichmentFuel rod diameterRange of temperature controlMaximum pressure on water annulusIrradiation in a reflector channel,

typical unperturbed thermalneutron flux

Outside diameter of test section

700nat.9. . .6015

1,5.43

. . 5%15

. .2,5.

W/cm

mm°Cbar

10 1 4 n/cm2.smm


Several integral control panels are available for followingthe operating conditions of the experiments concerning tempe-rature, fuel rod power, water filling and gas pressure in thedifferent leaktight pneumatic control circuits (cf .Figure 2.2.,scheme of irradiation device).

I S SAMPLING

WASTE

IN PILE SECTION

ROD POWER CALIBRATION PLUG

GAS GAP

RINSING TUBE

COOLING WATER

SS PRESSURE TUBE

SELFPOWERED NEUTRONDETECTOR

STAGNANT BOILING WATER

FUEL ROD

CROSS StCTION OF IN PILE PART

Fig 2.3.1 »WC (Pressurized Water Capsule)

11.

2.3. PWC

"Pressurized Water Capsule"

Principal uses :

Testing of prototype fuel rods for BWR and PWR application at realisticconditions concerning coolant pressure and temperature.

Design feature :

Cf. Figure 2.3.1.

The fuel rod is placed in a pressurized thimble tube filledwith demineralized water. The nuclear heat produced by thespecimen is radially dissipated through the stagnant boilingwater annulus at a constant temperature depending on thesaturation pressure. A small water flow is maintained throughthe inpile section for evacuating radiolytic gases and detectingcladding rupture by the presence of fission gases. Selfpoweredneutron detectors allow the registration of fuel rod powervariations. The rig may be put into a calorimetric devicefor determining the rod power and calibrating the selfpoweredneutron detectors. It is reloadable and allows the reirradiationof fuel rods.The water filling of the pressurized thimble tube can be eva-cuated and renewed. The rig may be charged into a He3 screenfor power cycling of the fuel rod (cf. "VNS").

Operating character is t ics :

Maximum fuel rod powerUsual fuel enrichmentFuel rod diameterActive fuel rod length, max.Coolant pressureCladding temperature

Irradiat ion in a reflector channel,typical unperturbed thermalneutron flux

Outside diameter of t es t section

Outpile control equiment :

600 to 700 W/cmnat. . . . 5%9 . . . 1 5 mm700 mm70 and 145 bar10 t o 20°C above b o i l i n gtemperature of coolan t

129

14 9.10 n/cnT s

mm

The available control installations can accept several irra-diation devices simultaneously. Coolant pressure, neutron fluxand water activity are registered ; a special safety systemacts in case of pressure loss in the inpile thimble tube,(cf. Figure 2.3.2.)

VENT.

GLOVE BOX

PRESSURIZEDNITROGEN

IN PILESECTION He 3 GAS GAP

SS INNER TUBE

IRRADIATION DEVICE

AI FILLING PLUG

COOLING WATER

SS OUTER TUBE

CROSS SECTION OF IN PILE PART

Fig.2.4 VN S (Variable Neutron Screen)

15.

2.4. VNS

"Variable Neutron Screen"

Principal uses :

Cycling of fuel rod power by varying the thermal neutron fluxincident on the specimen.

Design features :

Cf. Figure 2.4.

The annular gap between two concentric tubes contains Heliumgas highly enriched with the isotope He3 which possesses animportant absorption cross section for thermal neutrons. Byvarying the Helium gas pressure, the neutron flux in thecentral hole, occupied by a test section, can be changed.A small gas flow is maintained through the annular gap forpurifying the Helium from Tritium produced by neutron capture.The Helium pressure in the gap can be varied automaticallyaccording to the rhythm of power cycling specified for thecentral test section.


Range of He pressure variationThickness of annular He gapInside diameter of neutron screenRange of neutron flux depression

factorRate of power ramp,max. for short

range

1,5 .250 or

1 . . .

4

30

0 ,

3 6

3

a t ammmm

% /


The gas pressure in the annular gap of the inpile section isvaried by means of an outpile control system which allows toset-up frequency, rate and range of the power cycling wanted.The set point values are compared with a characteristicmeasurement taken from the test section (e.g. temperature ortemperature difference) .

17.

3. DEVICES FOR LIQUID METAL COOLED FAST BREEDER REACTOR (LMFBR)DEVELOPMENT

The irradiation testing of structural and fuel materials forsodium cooled fast reactors represents an important part ofthe experimental load of BR2.

Due to the high fast neutron flux in the central channelof the BR2 tubular fuel elements a fast neutron fluence up toabout 1,3 . IO22 n/cm2 (E >• 0,1 MeV) can be achieved in oneyear. The high fast flux allows also to produce in typical fastreactor fuel rods, heating rates by fast fission comparable tothose obtained in fast reactors. The irradiation devices used inthis case are surrounded by a tubular cadmium screen that reducesthe thermal neutron flux in the central experimental spacepractically to zero.

Besides those cadmium screened rigs, other irradiation deviceshave been developed where the thermal neutrons are admitted tothe specimens and are considered compatible with the irradia-tion objectives.

In most cases the irradiation samples are immersed in sodiumor in sodium-potassium eutectic alloy (NaK) for compatibilityand heat dissipation reasons.

In order to realise the elevated temperatures specified in generalfor the irradiation specimens, well defined thermal barriersare employed towards the BR2 cooling water being at about 50°C.

• * / • •

CA. GAS ANALYSERPT : GAS PURIFICATION TRAP

IN PILE SECTIONGMP [ TEMPERATURE CONTROL BY GAS MIXING)

BR2 FUEL ELEMENT

Cc^ SCREEN JOPnONAL)

ÇOOU_NG _WATER

SS ENVELOPING TUBE

REGULATING GAS GAP

SPECIMENS

NaK FILLING

SS CAPSULE TUBE

Fig.3.1


DISP (Device for Irradiation of Structural Parts)

V26

19.

3.1. DISP ;

"Device for' Irradiation of Structural Parts"

Principal uses :

Testing of structural material specimens immersed in sodium orNaK at elevated constant temperature and at high fast neutrondose.

Design features :

Cf. Figure 3.1.

The samples are contained in a leaktight stainless steel capsuletube filled with sodium or liquid NaK. A surrounding envelopingtube creates a concentric annular gas gap around the specimencapsule that serves as heat barrier for increasing the tempera-ture of the samples heated up mainly by absorption of gamma rays.The temperature of the specimens measured by thermocouples canbe regulated by varying the heat conduction through the gas gapeither by changing the gas composition (He/Ne) or by acting onon the gas pressure. The last system allows a fully automatictemperature control.The rigs may be equipped with an axial displacement mechanismfor shifting the test section in the neutron flux and with asurrounding cadmium screen for filtering the thermal neutrons.


Temperature of specimensUseful diameter of specimen capsuleActive length of test sectionUsual range of gamma heating rateIrradiation in a fuel element channel,

typical fast neutron flux(En > 0,1 MeV)

Outside diameter of test section

(* including Cd-screen)

25

20020 (7005 to

4 to

to22)

15

6.,4 or 3

700

i o 1 4

4*

°CmmmmW/g

n/cm .smm


Several integral control panels are available as well of the gasmixing type (GMP) as of the gas pressure regulating type (TCV)for controlling the operating conditions of the inpile sectionmainly concerning temperature, gas pressure, >;as composition,gas purity and leaktightness of the different containments andcircuits.The "TCV" panels possess a larger temperature regulationrange and allow to maintain automatically a given acr pointtemperature independent of fluctuations in the nuclear -.eatingrate of the samples, (cf. Figure 3.3.1.)

WASTE

VENT

CHARCOALTRAP

IN PILE SECTION

Be OR AI FILLING PLUG

COOLING WATER

SS ENVELOPING TUBE


Fig.3.2.i CIRCE (Calorimetric Irradiation Capsule Experiment.)

21.

3.2. CIRCE

"Calorimétric Irradiation Capsule Experiment"

Principal uses :

Irradiation of high performance fuel rods (mainly carbide fuel)at elevated cladding temperature in thermal neutron flux.Possibilities for elaborated inpile measurements.

Design features :

Cf. Figure 3.2.:.

A copper capsule of cruciform cross section contains in itscentral hole the fuel pins which are immersed in liquid metal(NaK). The copper capsule is surrounded by a stainless steelenveloping tube as barrier against the reactor cooling water,flowing outside. The thermal link between copper capsule andenveloping tube is realised by an extrusion process.At the inside wall/the copper capsule contains a layer of Nickelor a coextruded stainless steel tube as protection against thehot NaK.By means of thermocouple pairs in the four legs of the coppercross,the temperature gradients are measured and hence the fuelrod power. The cladding temperatures can be extrapolated fromthese measurements or directly registered by thermocouples at-tached to the cladding.The gas plenum of the copper capsule is connected to a sweepingcircuit for fission gas detection in case of fuel claddingrupture. Several fuel pins may be loaded into one capsule andby means of gas tagging,a ruptured pin can be identified.Besides the good calorimétric properties of the copper crosscapsule, it creates a considerable thermal neutron flux depres-sion in its central hole which favours the irradiation of highlyenriched fuel pins representative for fast reactor application.The basic design of the CIRCE device offers nummerous possibilitiesfor supplementary and particular instrumentation or operatingamendments.The "S" type version for instance was developed for measuringthe swelling of a fuel pin during irradiation : The liquid metalsurrounding the fuel rod and being displaced by swelling iscollected in a bellow chamber on top of the copper capsule.The supplementary volume pushed into the bellow chamber causesa corresponding vertical movement of a gauge rod linked to thebellow system which is precisely measured by an inductive linearmotion transducer.

• * / • •

xwxwwwwx

23 .

Figure 3.2.2. shows the swelling measurement device schematically.It allows :'.n addition to check the fission gas pressure developedby the fuel pin; this measurement is done by increasing thecontrol pressure in the upper bellow till equilibrium with thepressure in the lower bellow which communicates with the gasplenum of the fuel pin. A characteristic movement of the gaugerod indicates the pressure balance.Another version of the CIRCE device is the so called Fa?*- FluxCapsule ("FFC"). The outer enveloping tube of the copper crosssample holder is here replaced by a cadmium tube, sandwichedbetween two stainless steel tubes. It serves for the irradiationof fast reactor fuel pins in epithermal and fast neutron fluxwhile the thermal neutrons are screened off by the cadmium. Theoutside diameter of Lhe rig is 34 mm and it is loaded into thecentral hole of a 5-shell BR2 fuel element. Besides the cadmiumscreen which leads to a bigger diameter of the inpile section,all other design features are similar to the basic CIRCE rig.A further development of the CIRCE device concerns the possibi-lity for fuel rod power cycling by varying the fast neutronflux inside the Cadmium screen (PCF-Power Cycling in Fast flux).


Maximum fuel rod powerCladding temperatureFuel rod diameterUsual fuel enrichment (U235 or Pu239)Active fuel rod length, uptoOutside diameter of test sectionIrradiation in a reflector channel,

typical unperturbed thermalneutron flux

1200 .500.

6.15.

60029

. . 1500

. . 7009

. . 40

W/cm°Cmm%mmmm

1,5. . .2,5.10 1 4 n/cm2.s


Control panels are available for operating several CIRCE typedevices simultaneously. They are equipped with temperaturerecorders, pressure control circuits and alarm circuits. Theactivity of the sweeping gas is monitored and a glove boxinstallation allows sweeping gas sampling and analysing.The fuel rod swelling measurement of the CIRCE S type deviceis registered by a linear recorder.

CONNECTING HLADCONTROL

V

IN PILE SECTION

He

TCVITEMPERATURE CONTROLBY GAS PRESSURE VARIATION)


Fig.3.3.i CFC(Compatibility Fuel Cladding)

25.

3.3. CFC

"Compatibility Fuel-Cladding"

Principal uses :

Testing of fuel rods at well defined, constant cladding tempe-ratures; automatic temperature control.

Design features :

Cf. Figure 3.3.1.

The "CFC" capsule is a modified version of "CIRCE" describedabove. The main difference consists in a small gas gap whichis realised between the four wings of the copper cross andthe stainless steel enveloping tube. By means of well definedspacer pins attached to the cruciform copper capsule, it takesa concentric position in the surrounding tube. The gas gapsat the wings of the copper capsule (of the order of atenth of a mm thickness) create a heat transfer barrier forthe fission power developed by the fuel pins inside the coppercapsule. The heat conductivity of the gas gap (Helium) can bechanged by varying the gas pressure inside the enveloping tube.By means of the corresponding outpile control system the sampletemperature (e.g. cladding temperature) is maintained automa-tically at its set point value independent of variations innuclear heating rate of the specimens.The gas gaps between the wings of the copper capsule and theenveloping tube may have different thicknesses along the heightof the test section in order to realise a certain axial tempera-ture profile, e.g. same cladding temperature of a fuel pin at thegas plenum as at the fueled column.The usual operating gas pressures are ranging from about 10 to200 Torr. Several hundred degrees Celsius of temperature controlrange can be covered by the regulating system. Apart from thisparticularity of temperature regulation, the "CFC" rigs exhibitsimilar characteristics as the "CIRCE"devices, i.e. calorimetricheating rate measurement, cladding rupture monitoring, irradiationof highly enriched fuel and a considerable potential for sophisti-cated instrumentation.They can also be equipped with an axialdisplacement mechanism for shifting the test section in thereactor core.

27.


Maximum fuel rod power, about(depending on the specified claddingtemperature and control range)

Cladding temperature control range,usual

Fuel rod diameterUsual fuel enrichment(U235 or Pu239)Active test section length, up toOutside diameter of test sectionIrradiation in a reflector channel,typical unperturbed thermal neutronflux

500

450615

60029

... 700

... 7508

... 40

W/cm

°Cmm%mmmm

2.1014 n/cm2.s

V;


The "CFC" devices are operated by so-called "TCV" control panels(Temperature Control by Vacuum).Several of these panels existwhich provide the necessary Helium gas supply to the rig forregulating its pressure via a PID-temperature control.A vacuum pump creates a permanent depression at the gas outletof the rig; by means of an automatic feed valve the heliumsupply is controlled in function of a given set point temperaturewhich causes pressure increase in case of high temperature orpressure decrease for low temperature, (cf. Figure 3.3.2.)Apart from the temperature control system, the CFC devices canbe operated by means of the same outpile equipment as used forthe CIRCE rigs. It concerns mainly temperature registration,pressure (leaktightness) control, fission gas detection andgas sampling.

J

TEST SECTION

BOTTOM ENO

LONGITUDINAL SECTIONOF IN PILE PART

Be FILLING PLUG

COOLING WATER

S S _ ENVELOPINGTUBE

HEAT DIFFUSINGMATRIX

THERMOCOUPLES

PROT ECTING DISC

FUEL PF.

/CROSS SECTION OF IN PILE PART

Fig.34 POM (Pellets Of candidate fuel Material)

29.

3.4. POM

"Pellets Of candidate fuel Material"

Principal uses :

Irradiation of individual fuel pellets (discs) at very highspecific fission power for boosting fuel burn-up (parameter tests).Also compatibility tests between fuel and candidate claddingmaterials.

Design features :

Cf. Figure 3.4.

Several individual fuel pellets (up to 18) are separatelylocated in a SAP (Sintered Aluminium Powder) or in a Zircaloymatrix. Each pellet or disc occupies a matrix cavity which isclosed by a counter piece. The discs are staggered along theaxes of the test section in the way that their axes are per-pendicular to the first one. The heat produced in the fuel samplesis dissipated on both flats of the disc by solid contact withthe heat diffusing matrix which is hermetically enclosed by

a stainless steel tube. When the fuel discs are heated up, thenarrow gaps between sample and matrix and between the latterand the surrounding tube are filled up by thermal expansionof the matrix and a good solid contact between the differentparts of the capsule ensures the radial heat evacuation tothe reactor cooling water,flowing outside.Thermocouple pairs at different radii of the matrix allow tomeasure the heating rate of the samples. Between fuel discand matrix, samples of cladding material (also in form ofdisc) or a protecting foil may be located. Their temperaturecan be measured by incorporated thermocouples.As well oxide as carbide or other fuel compositions can be tested.


Specific fuel heating rate aboutDimensions of fuel discs, diameter

thicknessUsual fuel enrichmentUsual effective irradiation time

to reach 100.000 MWd/t burn-upOutside diameter of test sectionIrradiation in a reflector channel,typical unperturbed thermalneutron flux

12.000.. .15.000 W/cm3

9 or 10 mm1 , 5 mm

30...40 %

6029

80 daysmm

2,5. . .3,5 . 10 1 4 n/cm2.s

31.


Several control panels are available for operating the POM typerigs. They are equipped with temperature recorders, pressurecontrol circuits and alarm signalisation.Gas samples can be taken from the inpile section via a glovebox for detecting fission gases in case of leakage of the fuelcapsule.

. . / •

IN PILE SECTION

-XI—Ã-V8 MV42

<—xVIO MV40

/ PI

GAS SAMPLING INGLOVE BOX

V03

SV30

•VENT

CHARCOALTRAP

OUTER SS ENVELOPING TUBE

CAOMIUM SCREEN

THERMOCOUPLES ^

INNER SS ENVELOPING TUBE

WÊuàvs/x\ \ /

SS CAPSULE TUBE

NaK FILLING

FUEL PIN

HIGH TEMPERATURETHERMOCOUPLE


Figas FAFNIRIFuel Array Fast Neutron Irradiation Rig)

33.

3.5. FAFNIR

"Fuel Array Fast Neutron Irradiation Rig"

Principal uses:

Testing of fast reactor oxide fuel rods in fast neutronflux by means of a cadmium screen.

Design features:

Cf. Figure 3.5.

One or several fuel pins are contained in a NaK filledcapsule tube of stainless steel. The sample holder issurrounded by a cadmium screen sandwiched between twoconcentric stainless steel tubes. Between the fuel pincapsule and the cadmium screen exists a small gas gapconceived as a heat barrier for realizing elevated fuelcladding temperatures.The rig is equipped with thermocouples attached to thefuel pin cladding.The FAFNIR device type "C" contains pressure transducersconnected by capillary tubes to the fuel pin gas plenumfor measuring the evolution of fission gas pressurebuild-up.Another version of the rig type "D",allows the measurementof fuel pin centre temperature by means of high tempe-rature thermocouples.The temperature gradient across the 3 mm thick cadmiumscreen may be measured by incorporated thermocouple pairsfor a better evaluation of the fuel rod power.The fuel rods being completely shielded against thermalneutrons,show a flat fission rate distribution over thefuel cross section as typical for fast reactor conditions

Operating characteristics:

Maximum fuel rod powerCladding temperatureFuel rod diameterActive fuel rod length, up toUsual fuel enrichmentOutside diameter of test sectionCadmium screen lifetime, (nuclearburn-up) at full powerIrradiation in the central hole of a6 shell BR2 fuel element, typicalunperturbed fast neutron fluxabove 0,1 MeV

500. .600. .

6. .

250. .

.600

.700

.76009025

.300

W/cm°Cmmmm%mm

days

2,5.

35.

Outpile control equipment:

A number of FAPNIR control panels is available equippedwith temperature recorders, fission gas pressurerecorders, with pressurized gas circuits for leaktight-ness control and with alarm signalisation.Gas samples can be taken via a glove box for detectionof incidental fission gas liberation.

k • / • •

37,

3.6. FASOLD

Principal uses:

Testing of fast reactor high performance carbide fuelrods in fast neutron flux by means of a cadmium screen.

Design features:

Cf. Figure 3.6.

The fuel rods are contained in a Niobium capsule tubefilled with liquid Sodium-Potassium alloy (NaK).Niobiumhas been chosen as specimen containment in order to mini-mize the thermal stress in the capsule wall produced bythe elevated linear heating rate of the fuel rods.A stainless steel cladded cadmium tube surrounds thespecimen capsule for shielding against thermal neutrons.A small annular gap between cadmium screen and fuel rodcapsule allows for cooling water flovi. The cadmium screenof relatively small thickness (1,5 mm) can be replacedby a new one when it is consumed by nuclear transmutationduring the irradiation campaign.For a fuel burn-up value of 100.000 MWd/t the cadmiumscreen is renewed once or twice.The fuel rods can be instrumented by cladding thermocouples,centre thermocouples and pressure transducers for fissiongas pressure measurement.According to experience, cladding rupture of the fuel pinscan be detected by temperature oscillations of the thermo-couples attached to the cladding.


Maximum fuel rod powerCladding temperatureFuel rod diameterActive fuel rod length, up toUsual fuel enrichementOutside diameter of test sectionIrradiation in the central hole of <6 shell BR2 fuel element, typicalunperturbed fast neutron flux above0,1 MeV

1000. .500. .

7 . .6009025

. 1500

.700

.8

W/cm°Cmmmm%mm

4. . .6.1011* n/cm2.s


Several control panels are available for operating theFASOLD devices. They are equipped with temperature recor-ders, fission gas pressure recprders, with pressurizedgas circuits for leaktightness control and with alarm{signalisation.Sas samples can be taken via a glove box for detection ofincidental fission gas liberation.

?• I*.

^s

EM. PUMP UNIT

1

—I-

7% HOW METER

1HEAT EXCHANGER

EXPANSION TANK

TEST SECTION

IPSL 250kW

(IN PILE SODIUM LOOP)

GENERAL DISPOSITION

IN REACTOR CHANNEL

Fig. 3.7.1

39.

3.7. IN-PILE SODIUM LOOPS

3.7.1. In-pile sodium loop - 250 kW

Principal uses :

Testing of fuel rods for a LMFBR in an epithermal neutronspectrum, including the environment of sodium flow andtemperatures. The fuel pins are arranged in a bundle of 7pins .

Design features :

The loop consists of an integrated sodium circuit, an inter-mediate CO2 circuit and a water circuit. The primary, sodiumcircuit is placed inside a pressure tube and is located in a200 mm reactor channel. The whole sodium circuit, with pumpsand heat exchanger is included in one pressure tube and is alsohandled in the reactor as a compact unit. This is one of itsmain features.Other striking features in the design of the in-pile section,containing mainly the primary sodium circuit are :- The in-pile section is completely assembled, filled with

sodium and tested before loading in the reactor.- To ensure access to the reactor top cover and to facilitate

loading and unloading procedures of the in-pile section from theactivation shielding standpoint, the sodium in the upper partof the circuit can be drained into an expansion tank. Thetank is situated in the lower part of the in-pile sectionjust above the reactor core with its top cover fitting intoa heavy metal shield.

- A rupture of the cladding of the prototype fuel element isdetected by sweeping the cover gas and monitoring directlyfor Xe and Kr fission gases or by the gas tagging technique.

- For dismantling, the in-pile section can be transferred ,under water, from the reactor pool through the storage canaltowards the high-activity hot cell connected to the reactorcontainment building.

- An external concentric cadmium screen around the test finger,cooled by the reactor water, is used to filter out the thermalcomponent of the reactor neutron flux.

The sodium flowrate is adjusted in the system by changing thevoltage of the electromagnetic pumps. The temperature levelin the bundle region is adjusted by changing the CO2 flowratein the Na-CO2 heat exchanger.Figure 3.7.1. shows how the loop is installed in the reactorvessel.

i '

COOLING _ L ELECTRIC

WATER ( V I H E A T E R

Na COOLANT

FUEL RODSUB ASSEMBLY \ NI-DELAYED NEUTRON DETECTIONJ

/ARI-FISSION GAS DETECTION

TEST SECTION CROSS SECTION ATREACTOR CORE MID PLANE

Fig 17 2 IPSL - 500 kW

J

41.


- Fuel pin

- fuel rod power. Pu-U mixed oxide. Pu-U mixed carbide

- fuel pin diameteractive fissile length- total length- number of parallel pins oxide

carbide

50010006. .75001080

73

W.cm *W.cm"1

mmmmmm

- Loop design

The design conditions for the primary circuit are as follows

- Maximum working temperature- Maximum working pressure at 600°C- Flow and pressure head

. for one pump at 100% output

. for both pumps at 70% output

- Total power output- Material of circuit stainless steel AISI 316- Sodium volume 10 &

60015

1111

250

.07

.5

°Ckg

&.kgI.kgkW

• cm~2

s-1— 2. cm-*

s-1.cm"2

Outpile equipment

The outpile equipment consists of the heat evacuation system :the secondary CO2 gas circuit and the water circuit. Thedifferent loop parameters are measured, controlled and regu-lated from the loop control panel. The parameters necessaryfor the calculation of the loop power, linear power, andsurface temperature are measured by a small computer, printingout the important parameters of the loop such as linear powersand cladding surface temperatures.

(7) FUEL ROD SUB ASSEMBLY

(2) SODIUM COOLANT

(3) THERMAL INSULATION

(Z) PRESSURE TUBE

(5) STAGNANT He GAP

(6) ENVELOPE TUBE

CADMIUM SCREEN

Al FILLING PLUG

(9J DRIVER FUEL ELEMENT

(ÍO) REACTOR CHENAL $200

(Tf) COOLING WATER

CROSS SECTION OF INPILE PART

Fig 373 IPSL-500kW

43.

3.7.2. In-pile sodium loop - 500 kw

Principal uses :

Testing of fuel rods for a LMFBR in an epithermal neutron spectrum,including the evironment of sodium flow and temperatures.The fuel pins are arranged in a sub-assembly of maximum 19 rods.

Design features :

The design of the loop is similar to the in-pile sodium loop 250 kW(see 3.7.1.) but with modified heat evacuation circuits to evacuate500 kW instead of 250 kw. Larger sodium pumps and heat exchanger areprovided for this purpose, together with a bigger experimental cavity.The secondary heat evacuation fluid is helium, which operates at apressure of 30 kg.cur2. The circuit contains three gasbearing circulators.

Figure 3.7.3. shows a cross section through the in-pile part with the fuelrod bundle, the loop containment tubes, the cadmium screen and the BR2driver fuel elements.The design features of the IPSL device allow its use for different experi-mental objectives going from endurance tests (accumulation of fuel burn-up)to more sophisticated research on fuel assembly behaviour as in case ofsodium coolant flow blockage ("MOL 7C" project).


5006..7108019

W.cmmmmm

,-1

- Fuel pins

- fuel rod power (mixed oxide)- fuel pin diameteractive fissile length- number of pins

Loop design

The main characteristics of the primary circuit are as follows:

fluidmaximum sodium temperatureoperating pressuredesign pressureflow rateheat to be evacuatedsodium volume

sodium7006402.5

50019

°Ckg.cm"2

kg.cm-2

Jl.S"1

kWi.

Outpile equipment :

The equipment outpile contains the components and piping for theheat evacuation systems. The loop parameters are measured, controlledand regulated from the main loop control panel.The out-of-pile equipment is the same for all irradiation campaigns,contrary to the in-pile section which is new for every new irradiationof a fuel bundle.

Plate

DCOcs

so

õ

2EX4Î

45.

4. DEVICES FOR GAS COOLED BREEDER REACTOR (GBR)DEVELOPMENT

4.1. "GSB" LOOP

Principal uses :

Performance testing of GBR-type vented fuel pins arranged ina bundle of 12 pins under an epithermal neutron flux anddynamic cooling conditions similar as in a GBR.

• -N.

Design features :

Except the smaller number of fuel rodo, 12 in the present case,and the reduced rod length (600 mm active length) , the test fuelelement represents the present design of the vented fuel bundlein its technical details. The irradiation is performed in acadmium shielded in-pile section. The system pressure of the loopis 60 kg.cm"2;helium coolant flow and temperature capabilitiescover the range of the operating conditions a fuel element mayundergo in the GBR. The pressure equalization and fission gasventing system used in experiments is equivalent to that offuture gas breeder reactors, (cf. Figure 4.2.)The loop facility consists mainly of the following subassemblies :main loop, in-pile section, fuel element transfer device, heliumpurification fission gas separator system and the secondaryequipments. The relative arrangement is shown in figure 4.1.There is a particular emphasis in the design on the requiredcooling of the test fuel element under all conditions.


- Fuel pins- fuel rod power - Pu-U mixed oxide- fuel rod diameteractive fissile length- total length- number of pins

Loop design data :- cooling gas helium- operating pressure 61- mass flow (downward through bundle) 0,25- total power 300

5 0 07 , 0

6 0 09 2 0

12

W.cmmmmmmm

-1

kg .cmkg.s"1

kw

-2

Outpile equipment :

The whole loop is installed outside the reactor vessel, exceptthe in-pile section which passes through the reactor vessel fromthe top to the bottom. The test fuel element is unloaded at thetop, while the two cooling gas connections are made at the bottom,Practically the whole circuit is installed in a shielded contain-ment, which has also the function of a ventilated double contain-ment .

Ci)©

©(s)

FUEL ROO SUB ASSEMBLY

He GAS COOLANT

GUIDING TU8E

PRIMARY CONTAINMENTTUBE (1.0 75mm)

STAGNANT He GAS GAP

© ENVELOPING PRESSURETUBE (O.D 97mm)

(T) CADMIUM SCREEN [O.D. 110mm)

(8) DRIVER FUEL ELEMENT

(?) REACTOR CHANNEL $ 200

()0) COOLING VK4TER

CROSS SECTION OF INPILE PART

Fig 4 2 GSB-LOOP

47.

5. DEVICES FOR HIGH TEMPERATURE GAS COOLED REACTOR (HTR)DEVELOPMENT.

Since the early years of operation,BR2 reactor was exten-sively used for testing structural and fuel materialsconceived for application in the core of gas cooled reactors.At the beginning, in the years 1963 to 1969, emphasis waslaid on the investigation of the behaviour of graphitesamples under intensive gamma ray and neutron exposure.The tests aimed in particular at the development of theAdvanced Gas cooled Reactor (AGR) and the high temperatureDRAGON reactor. Many irradiation rigs as well as a Heliumloop and a C02~l°op were loaded into BR2 for studying thephysical and chemical changes undergone by different kinds ofgraphite in a nuclear radiation field. The high fast neutronflux (up to 5.1011* n/cm2s. above 0,1 MeV) and the intensivegamma ray exposure (up to 18 W/g in graphite) availablein the central hole of BR2 fuel elements were particularattractive to simulate within a few months the effectsof years of power reactor operation.

Sufficient data on the nuclear behaviour of graphitescould be accumulated during these years and the specificirradiation devices developed for this program ceasedoperation. They are therefore not described here.

Another important irradiation program however emerged sinceend of the sixties and is continuing with increasing effort,it is the irradiation testing of coated fuel particles. Thiswork is devoted to the development of fuel materials forgraphite moderated reactors with high cooling gas temperature(HTR line). The irradiation conditions offered in the BR2core are especially suited for investigating the behaviour ofcoated fuel particles because of the flexibility in choosingirradiation channels which allow a realistic simulation ofirradiation history with respect to fast neutron dose, fuelburn-up, gamma ray heating and fuel temperature evolution.

Based on earlier experience, a new type of irradiation rig hasbeen developed, called "CPR", which is now currently used.It is described below.

Also project studies were performed towards a loop experimentfor testing large samples of HTR block fuel elements underdynamic Helium flow cooling conditions.

• • / • •

TOP END OFTEST-SECTION

SS ENVELOPING TUBE

SS CAPSULE TUBE


Fig51 CPR (Coated Particles Rig )

49.

5.1. C P R

"Coated Particles Rig"

Principal uses:

Irradiation of coated fuel particles contained in agraphite matrix; measurement of fission product releaserate; effect of fuel power and temperature changes.

Design features:

Cf. Figure 5.1.

The coated fuel particles are contained in a cylindricalgraphite matrix being enclosed by a stainless steelcapsule. The matrix array may differ according to theparticle distribution and the fuel composition chosen. Asecond stainless steel enveloping tube surrounds in generalseveral specimen capsules which are axially superposed.It is cooled at the outside surface by the primary reactorwater. An annular gas gap between the enveloping tube and thefuel capsules serves as heat flux barrier and as entry ductfor the sweeping gas which is used simultaneously fortemperature regulation of the fuel specimens by varyingthe ratio of the Helium/Neon gas mixture. The sweeping gaspasses through the fuel capsules and carries fission productsliberated from the fuel particles to the outpile samplingand analysing system. A second annular gas gap exists ingeneral between the fuel matrix and the fuel capsule forincreasing the temperature gradient towards the specimens.

High temperature thermocouples are mounted to the graphitematrix for measuring and regulating the fuel temperature.

The CPR devices are usually equipped with an axial translationmechanism for following the shift of neutron flux peakduring a reactor operating cycle.

A special computer program elaborated for the CPR experimentsallows to estimate fuel temperature, fuel power and burn-upevolution, fuel breeding build-up, fast neutron dose andfission product release rate. The latter being expressed asratio of fission product release rate over birth rate.

The CPR devices are irradiated in the central hole of BR2fuel elements. During their irradiation campaign they aretransferred into different positions of the reactor corefor simulating an irradiation history representative forHTR fuel.

51 .

Different c"\ameters of inpile test section can be realisedaccording to the sample size to be loaded. The "C" type rigshave an outside diameter of 25 mm, the "E" type amountsto 34 mm and the "G" type show an outside diameter of 79 mmfor testing fuel rods in a block element array.

A thermal neutron screen may be put around the rigs forlimiting the fuel particles power output at the beginning ofirradiation campaign while boosting the fast neutron exposure.

The nuclear power production in the inpile section is ingeneral not limited by the heat transfer characteristicsof the rig but by the maximum values specified for thefuel particles. The fuel loading consists usually of amixture of feed (Uranium) and breed (Thorium) particles.


Maximum heat production rateas specified for a fuel particleof 200 ym kernel diameterFuel enrichment (U 235)Temperature of coated fuel particles,usually specified range ;Burn-up of feed particlesFast neutron dose (above 0,1 MeV)Number of fuel capsules per rigActive length of fuel charge, up toOutside diameter of test section,"C" type, loaded in a 6 shell BR2 fuel element"E" type, loaded in a 5 shell BR2 fuel element"G" type, loaded in a standard core channelTypical unperturbed fast neutronflux above 0,1 MeV 2.Typical unperturbed thermal neutro/» flux 1,5.

0.

9.2.

ntt

0,290

. .140075%

W%

°cFIMA

..10.1021 n/cm2

. .4600

253479

mm

mmmmmm

, 4.10ltt n/cm2s, 3.1011* n/cm2s


Several control units are available at BR2 which allowsimultaneous irradiation of a certain number of CPR devices.Apart from temperature and gas pressure measurementequipment, the outpile control installations are mainlycharacterized by a fission product sweeping and analyzingcircuit and by a temperature regulating system based onHelium/Neon gas mixture variation.The outpile control equipment is shown on Figure 5.2.

53.

6. DEVICES FOR ISOTOPE PRODUCTION

Due to the high thermal and epithermal neutron fluxavailable in BR2, an important routine production ofradioisotopes was established and continues with increasinginterest. It concerns mainly the production of Cobalt 60,of Iridium 192 and of numerous small samples of all kindsof elements whose radioactive isotopes are of interest forresearch, process and medical application. The elevated neutronflux makes possible to obtain outstanding values of specificactivity which amount to 400 Ci/g for Co 60 and 700 Ci/gfor Ir 192 for instance.

The isotope production at BR2 is performed by means of largelystandardized irradiation capsules and devices, which arebriefly described below.

• • / • •

SUSPENSIONTUBE

BAIONETTECOUPLING

IRRADIATIONBASKET

V+771

COOLINGSLOT

\

LOCKINGROD

TOP END

LONGITUDINAL SECTION

BASKET TUBE

COOLING WATER


OPEN BASKET ( R A - T y p e )

V

FOOT END

V-771

Be FILLING PLUG

IRRADIATION CAPSULE

" "N.

55.

'"•s.

6.1. OPEN BASKETS (R-TYPE)

Two kinds of so-called open baskets exist. They arereloadable with isotope capsules. One is used in reflec-tor channels, the other in the central channel of 5-shellfuel elements.

Their characteristics are :

In reflector position- Denomination RA/25- Outside diameter 29,5 mm- useful inner diameter 25 mm- Useful inner length 900 mm- Maximum length of an individualcapsule to be charged into the basket 100 mm

- Cf. Figure 6.1.1.

In fuel element position- Denomination RE/25- Outside diameter 34 mm- Useful inner diameter 25 mm- Useful inner length 900 mm- Maximum length of an individualcapsule to be charged into thebasket, up to 900 mm

- Cf. Figure 6.1,2.

The baskets consist of an aluminium tube fitted at bothends with adaption pieces. Loading and unloading ofirradiation capsules is realised via the top end andusually under water in the reactor pool. The chargingopening is locked by mounting the suspension tube to thebasket.

The reflector baskets are provided with lateral aperturesfor cooling water access to the capsules.

For the fuel element channels, the basket tube has noapertures but four longitudinal fins at the inner wall forcentering the irradiation capsules and allowing thecooling water flow passing inside the basket betweencapsules and tube wall.

The admissible heat flux at the capsules surface is ingeneral limited to about 50 to 100 W/cm2. Higher heatingrates may be accepted but need a special examination.

ih

SUSPENSIONTUBE

COOLINGWATERINLET

BAIONETTECOUPLING

IRRADIATIONBASKET

W771

V-72O

FUEL ELEMENT

ÇOOUNG WATER

FOOT END

7-461

LONGITUDINAL SECTION

BASKET TUBE

IRRADIATION CAPSULE


Fig.M.i OPEN BASKET I RE-Type)

57,

6.2. ISOTOPE CAPSULES

Two types of standardized capsules are generally usedtogether with the Open Baskets described above forirradiating small quantities of a large variety of elementsand compounds.

The capsules are made of aluminium and are hermeticallyclosed by welded covers. They are shown on figure 6.2.

'C S F" capsule

This irradiation capsule provided with a cold welded coverand an inner graphite cylinder with bore holes (sample holder),is used on routine base for loading samples into the reactorwhich consist of approved materials concerning chemicalcomposition and quantity. Its use is limited to specimensof relative low heat production and low induced activityas it is the case for the elements and compounds comprised onthe list of standard irradiations. Also concerning durationof irradiation, the "CSF" capsules are limited to a maximumof a few months. Their advantage lies in the simple andcommercialized method of encapsulation.

When used in the R-type Open Baskets, the "CSF" capsulesare surrounded by a perforated adapting sleeve of aluminiumfor fitting the basket hole.

Outside dimensions: diameter 22 mm, length 69 mm.

"I R" capsule

For the irradiation of small samples which represent withrespect to quantity,induced activity, heat production,chemical composition, physical characteristics orirradiation time ,essential differences compared to theapproved standardized specimens, the "IR" capsules areused. They consist of a cylindrical Aluminium blockprovided with bore holes of variable diameter accordingto the dimensions of the specimens to be loaded. The capsulesare hermetically closed by ah- electron-beam welded coverwhich possesses a tubular appendix for realising an appropriategas filling of the capsule. Because of its good heat conduc-tivity and inertness, Helium gas is usually employed forfilling.

The "IR" capsules are regularly used for the production ofradioactive Iridium sources.

Outside dimensions: diameter 25 mmlength 50 or 100 mm.

*=?

. , ' y / / / / /

///y/.-

'"-//•////////s

52

69

W,' / //y

/ / / / /

\7,

Y/

SPECIMEN

GRAPHITE CYLINDER

CSF CAPSULE

NEUTRON DOSE DETECTOR

• /

. / / / . ' / / / / / / /

82J32J

100 (50)

Fig.62 R CAPSULE

59.

Irradiation samples in form of powder which is frequentlyused, are enclosed in phiales of quartz for loading into"CSF" or "IR" capsules. The powder must be anhydrous andresistant against thermal decomposition and radiolysis.

IRRADIATION BASKET "RG 24'

CAPSULE FOR COBALT DISCS

Figeai CAPSULE FOR COBALT PELLETS

61.

6.3. COBALT IRRADIATION DEVICES

Several devices are used at BR2 reactor for producingradioactive cobalt 60 sources. The samples are in formof discs (0 20x1 mm or 0 10x1 mm) or of small pellets(0 lxl mm).

The cobalt discs are irradiated in aluminium capsules oftriangular cross section as shown on figure 6.3.1..These electron-beam welded capsules have a length of 100 mmand are loaded into irradiation baskets with a speciallyextruded aluminium profile ("RG 24" baskets) which fittsthe standard 33 mm diameter reflector channels. Eachcapsule contains 12 cobalt discs of 20 mm diameter; thealuminium cladding is collapsed on the discs by externalpressure for improving heat evacuation. One "RG 24" basket,contains in general 9 triangular cobalt capsules.

The cobalt pellets, on the other hand, are irradiated- in anannular array in individual cylindrical helium filled capsulesas shown on figure 6.3.1. or in special sandwiched tubesattached to the foot end of the BR2 control rods (cf.fiçure- 6.3.2.). The cylindrical capsules are loaded intothe "RA 25" or "RE 25" standard irradiation baskets andcontain about 65 g of cobalt each. The cobalt charge of thecontrol rod foot end devices amounts to about 180 g.

After irradiation the cobalt specimens are recovered in theBR2 hot cell by dissolving the aluminium cladding in-caustic solution or by mechanical dismantling if possible.A nickel layer on the cobalt surface prevents corrosion andcontamination.

According to irradiation time and irradiation position thespecific activity of the cobalt sources ranges usually between150 and 300 Ci/g. The annual production of Co 60 at BR2reactor is actually of the order of 200.000 Ci.

—s.

CHANNEL MATRIX (Be)

COOLING WATER

CONTROL ROD GUIDING TUBE1AI)

OUTER CLADDING TUBE ( A i l

Fig.632 COBALT CAPSULE AT CONTROL ROD FOOT END

63,

6.4. THIMBLE TUBES ("DG- 36" AND "DGR")

The BR2 reactor core is contained in a pressurized vesselbeing hermetically closed during operation. Loading andunloading of experimental devices is in general onlypossible during the shut-down periods for refueling.By means of so-called thimble tubes , whichpenetrate from the reactor pool into the core, someirradiation channels are accessible during operation forrealising short term irradiations indépendant of the reactorcycles.

The thimble tubes made of aluminium, communicate with thereactor pool and are filled with stagnant water. They havean inner diameter of 36 mm (outside 42 mm) and are usuallycharged with open baskets of the "RA 25" type containingstandardized isotope capsules.

The use of the thimble tubes is limited by the relativelypoor heat dissipation capacity from the samples through thestagnant water gap to the tube wall which is cooled at theoutside by the primary reactor cooling water. An upper limitof only 25 W/cm2 can be tolerated as heat flux on the surfaceof irradiation capsules being charged into the "DG 36".Because of the risk of pool water contamination in case ofcapsule rupture, only samples of low activity hazard areaccepted.

In general the thimble tubes occupy irradiation channels inthe outer reflector region of BR2 with a relatively lowgamma heating rate and a moderate thermal neutron flux.They are surrounded by an aluminium obturating sleeve tofit an irradiation channel of 84 mm diameter.

For improving the heat evacuation capacity of the thimbletube type irradiations a new device is actually underconstruction, called "DGR", which provides the inpile sectionwith an individual closed circuit of forced water cooling.It combines the advantage of on-load charge of irradiationrigs (also instrumented test sections) with the possibilityof accepting elevated nuclear heat ratings (up to 100 kW)and samples with increased contamination hazard.

WATER SUPPLY

RECIRCULATINGWATER TRACK

REACTOR

VESSEL

SAMPLE DISCHARGE

SCHEMATIC FLOW SHEET

H.R. (Hydraulic Rabbit)

SAMPLE CARRIER

Fig.6.5i H.R. ( Hydraulic Rabbit )

65.

6.5. HYDRAULIC RABBIT (HR)

Short term irradiations of small individual capsules areexecuted in a closed water circuit called Hydraulic Rabbit("HR"). The flow of demineralized water assures simultaneouslythe cooling of the samples and the injection and return (byflow inversion) of the irradiation capsules. They are exposedto neutron flux in a thimble tube plunging into the reactorcore. Charging and discharging of samples is realised outsidethe reactor containment building in a water pool. Two pipesconnect the terminal station with the inpile section: one,to convey the irradiation capsules to and fro (insidediameter 26 mm) and the other for recirculating the loopwater.

Tho main characteristics of the Hydraulic Rabbit aresummarized hereafter :

Water pressure, nominalWater flow-rate, nominalWater entry temperature, aboutHeat evacuation capacityMaximum admissible heat fluxat surface of irradiationcapsule, nominal

exceptionalAdmissible dimensions ofirradiatior capsules,when used together with astandard capsule carrier, diameter

length

when used without carrier,diameter of central partdiameter of guiding endslength

/cm'

250.

15 kg5 m3/h

40 °C34 kW

150 W/cm2

,300 W/cm2

16 ram150 mm

2 0 mm2 4 mm

214 mm

The standard isotope capsules, type "CSF" are acceptedwithout carrier.

The "HR" inpile section can be charged with up to 3 capsulesof maximum length of 214 mm each. It is an aluminiumthimble tube with re-entry water flow at the top end andan outside diameter of 45 mm. By means of a channel adaptingsleeve it is located in a standard reflector positionof 84 mm diameter and of a peak thermal neutron flux ofabout 2 to 3.1014 n/cm2.s.

As well isotope capsules as special capsules (rods) containingfuel (uranium and/or Plutonium) can be charged. Highperformance capsules, like fuel rods, are not allowed howeverto be withdrawn from the core at full reactor powerfor avoiding burn-out of the specimen during inversion ofthe coolant flow for recalling the rabbit. The discharge ofthese capsules demands a reduction of reactor power or

67.

waiting for the end of reactor operation cycle.

The position of the irradiation sample relative to the coremid-plane (hot plane) can be chosen by means of preinjectedspacer tubes of adequate length.

Because of the regular and continuous use of the HydraulicRabbit for isotope production, the irradiation time persample is limited to a maximum of a few days.

By injecting activation probes before or after an irradiationthe neutron flux can be exactly measured.

Figure 6.5.1. shows a schematic flow-sheet of the "HR"and gives the dimensions of the usual sample carrier.

6.6. POOL SIDE TUBES

Certain irradiations can also be realized outside the reactorpressure vessel by means of 4 so-called Pool Side Tubes (ID86 mm). These are inclined tubes descending in the reactorpool closely to the vessel wall. They can be charged withisotope capsules (mainly "CSF" type) contained in a cylindricalaluminium barrel (diameter 74 mm) with four bore hole positions(diameter 23 mm, height 70 mm).The loading and unloading of samples in the Pool Side Tubesis executed under water from the top of the reactor pool bymeans of a manually operated mechanism.The available thermal neutron flux ranges between 0,5.10J

and 3.1013 n/cm2.s . It shows a relatively steep gradientin axial direction of the Pool Side Tubes.

,13

EXPLODED VIEW OF BR2 REACTOR

VESSEL AND POOL

Fig. 7.1

69.

7. OPERATING AND IRRADIATION CHARACTERISTICS OF BR2

The BR2 is a high flux engineering test reactor which differs fromcomparable material testing reactors by its specific corearray.

The core is composed of hexagonal beryllium blocks with centralchannels. These channels form a twisted hyperbolic bundle whichis close together at the midplane but more apart at the lowerand upper ends where the channels penetrate through the coversof the reactor pressure vessel. By this array,a high fuel densityis achieved in the middle part of the vessel (reactor core) whileleaving enough space at the outends for easy access to the channelopenings (cf. figure 7.1.).

The BR2 fuel elements consist of several concentric tubular shells(up to 6) of uranium- aluminium alloy cladded by aluminium whichprovide a central channel for locating irradiation experiments.Besides these fuel element channels which offer a particularhigh fast neutron flux (up to 7.10*4 n/cm^.s), a large numberof channels exists in the beryllium matrix where no fuel elementsare loaded, (cf. Figure 7.2. and Figure 7.3.)

These reflector channels can be occupied by experiments whichdemand only thermal neutron flux or they are obturated by berylliumfilling plugs.

With an adequate core configuration a peak thermal neutron fluxof almost 1.1015 n/cm2.s can be achieved in the central reflectorisland of the BR2 core matrix.

The large number of identical channels in the beryllium matrixprovides the possibility for a great variety of different coreconfigurations depending on the demands of the experimental load,(cf. Figure 7.4.)By the use of fuel elements with incorporated burnable poison(boron and samarium), a considerable amount of negative reactivitycaused by strongly neutron absorbing experiments can be acceptedwhile realizing still a sufficient long reactor operating cycle(20 to 25 days). Up to 10 # of negative reactivity have been handledalready with certain experimental loads.

The actual reactor power varies with the core configuration beingused and may attain a maximum of about 100 to 120 MWth. For thepresent time the reactor is operated at around 80 MWth.

J

BR2 Be FILLING PLUG ( Cross section)

EXPERIMENTAL DEVICEOR FILLING PLUG

F,g 7.2 BR 2 FUEL ELEMENT ( Cross section )

73.

The BR2 reactor is cooled by light water flowing downward inthe Beryllium matrix. The cooling water represents together withthe hexagonal Beryllium blocks the neutron moderator. A constantpressure drop is maintained over the height of the core matrix(about 2,8 kg/cm2)which creates cooling water flow rates in thedifferent channels corresponding to the flow sections madeavailable.

The coolant flow velocity in the annular sections between theBR2 fuel element plates is about 10 m/s. It allows a specificheat flux of 450 to 500 W/cm2 at the fuel plates' surface (hotplane) without reaching surface boiling.During particular tests a peak value of 600 W/cm2 was even realized.

The pressure of the primary cooling water is 12,6kg/cm2at the entry ofthe reactor vessel. Its entry temperature ranges between 40 and50°C depending on the recooling conditions of the secondarycircuit which passes through four cooling towers.

A water pool houses the reactor vessel and gives the necessarybiologic shielding when operating the reactor or when chargingand discharging active material. This is done via the top cover ofthe reactor vessel by unlocking the channel shut-off plugsi Alsothe lower cover of the reactor vessel is accessible during shut-down periods by passing through the shielded sub-pile room.Some standard irradiation channels (84 mm diameter) and all ofthe 5 big 200 mm diameter channels have also shut-off plugs atthe lower cover for connecting experimental devices.

Around the reactor pool, on 6 different floor levels sufficientspace can be made available for installing outpile controlequipment of irradiation experiments. A transfer chute in thereactor pool creates the under water link with the BR2 dismantlinghot cell (60.000 Ci - 2,75 Mev gammas) for recovering anddispatching the irradiated samples or test sections.

A number of beam tubes (300 mm diameter) directed radially (5)or tangentially (4) to the reactor vessel at core level allowsntitron beam experiments with high yields.

By means of a neutron radiography device installed in the reactorpool, non-destructive examination of test sections before,duringand after their irradiation campaign can be executed. Shots canbe made as well with thermal as with epithermal neutrons.

An electronic data collecting system with 400 connections allowsthe registration and surveillance of measurements furnished bythe irradiation experiments.

The central cavity of burnt BR2 fuel elements can be used forexperiments under intensive gamma ray exposure (several times10' rad/h).

J

77,

gl

'•f's

A full scale core mock-up facility, called BR O2 is availablefor studying the neutron flux pattern over the reactor core infunction of modifications in the core configuration.Its Beryllium matrix represents an exact model of the BR2 coreand can be charged with the same fuel elements. The necessarypower for studying neutron flux distribution and core configurationproblems is about 100 W at maximum when using fresh fuel.Partially burnt fuel elements demand a somewhat higher power level.

The core mock-up facility is installed in a water pool betweenreactor containment building and dismantling hot cell.

A summary of the most interesting operating and irradiationcharacteristics of BR2 is given in the following.

Cooling system :

- Maximum heat dissipation capacity- Primary cooling water flow-rate- Primary cooling water volume- Flow-rate of primary coolant by-pass

purification system- Shut-down coolant flow-rate- Nominal cooling water pressure at

entry of reactor pressure vessel- Coolant entry temperature 40- Direction of cooling water flow in

the core- Coolant pressure difference across

the core- Coolant velocity between the fuel

plates- Maximum heat flux at the surface

of the fuel plates 450- 4 primary coolant pumps (3 operating,

1 in stand-by)- 3 intermediate (primary-secondary)

heat exchangers- 1 electrical heated pressurizing vessel

(197°C, 14 bar) serving as well fordegassification and as expansion volume

- Secondary cooling water flow-rate- 5 induced draft cooling towers

1207000150

30800

12,6. .50

downward

2,8

10,7

. .500

MWm3/hm3

m3/hm3/h

kg/cm2

°C

kg/cm^

m/s

W/cm2

7000 m3/h

1

-300 -200 -100 0 100 200

BR2 FUEL CHANNEL

300 mm

L- - t -

f—h-+-4-- --

i

, ...L.-i . g—.: 8

K|!

START OF CYCLE

-300 -200 -100 0 100 200

BR2 REFLECTOR CHANNEL

300 mm

TYPICAL AXIAL THERMAL NEUTRON FLUX DISTRIBUTION

Fig.7.8

79.

Ventilation system :

- Air flow-rate through BR2 containmentbuilding (4 changes per hour, partlyrecirculated)

- Depression in containment building- Normal flow-rate through stack of

non-recycled and vented air ofBR2 complex

- Air flow-rate through stack whenair recirculation being interrupted

- Depression in non-recycling airsystem (max.)

- Air flow-rate through iodine scrubberin case of emergency

- Height of stack above ground- Air exit velocity at stack with

maximum flow-rate (475.000 m3/h)

85.00010. .15

120.000

245.000

30

8.00060

10

m3/hmm WG

m3/h

m3/h

mm WG

m3/hm

m/s

Nuclear Data :

Number of irradiation positions (channels)in the Beryllium core matrix

with 84,2 mm diameterwith 50 mm diameterwith 203,4 mm diameter

Total

Concentric shell type fuel elementsfuel enrichment (U 235)fuel content (U 235) of a full size

element (6 shells)U-Al alloy type elementCermet type element with burnablepoison

height of fueled matrix

Number of fuel elements loaded : variableaccording to experimental charge,

minimum, aboutat present (configuration n°8), about

Maximum unperturbed thermal neutron fluxthat can be made available with an adequatecore configuration

Usual range of available unperturbed peakthermal neutron flux

Maximum unperturbed peak fast neutron fluxabove 0,1 MeV available in the central holeof a fuel element

64105

79

90%

242

400760

1536

gmm

0,9 to 1.1015

1,2 to 6.1014

n, c m . s

2n/cm .s

7.IO14 n/cm2.s

80.

- Usual range of available unperturbeõpeak fast neutron flux above 0,1 MeV

- Ratio of peak neutron flux to r.ieanneutron flux over the svis of anirradiation channel (760 mm length), about

- Heating rate by absorption of gammarays in materials,

in fuel element channelsin reflector channels

- Useful diameter of irradiation channels

3 to 6.1014 n/cm2!

1,55

5 to 16 W/g Al1 to 6 W/g Al

in' fuel elements: typetypetypetypetypetypetypetype

in reflector: centralcentral

central

6 i6 n5 i5 n4 i4 n3 i3 n

hole ofhole of

hole of

big channels

Note: By means of special plugs

Be-plugBe-hexagon(standard)Be-hexagon(small)

17,4262634,534,5434351 ,629,5

80

46200

mmmmmmmmmmmmmmmmmm

mm

mmmm

(e.g. of Aluminium)diate experimental diameters can be made available inthe 3 kinds of Be-matrix channels.Also special fuel elements can be arranged in the bigchannels (200 mm) around a test section for boostingthe fast neutron flux.

- Height of Beryllium matrix

- Diameter of Beryllium matrix

- Pitch of standard matrix channels,at core midplaneat reactor vessel cover

- Inclination of channel axis againstvertical, maximum at core periphery

at core center line

- Twisting angle between core midplaneand reactor vessel cover

- Number of beam ports,radial to core pressure vesseltangential to core pressure vessel

- Diameter of beam tubes

- Available thermal neutron fluxesfor beam port experiments (depending oncore configuration) ,

near reactor vesselat distance of beam ports

914

1054

96221

10°0°

mm

mm

mmmm

24"

56°44 '

58

300 mm

1 . .5. 10 1 3 n/cm?s1..5.108 n/cm?s

J

81

/ ; •

Acknowledgement.

The authors wish to express their gratitude to Mr.J. Swinnen for his contribution in preparing the figurescontained in this report.

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5ATING CHARACTERISTICS CEN/SCK - IAEA