CA14,0261?

AECL-5307

ATOMIC ENERGY # 9 ? ^ L'ENERGIE ATOMIQUEOF CANADA LIMITED ^ £ 9 DU CANADA LIMITEE

LATTICE MEASUREMENTS WITH 37-ELEMENT BRUCE

REACTOR FUEL IN HEAVY WATER MODERATOR:

DETAILED LATTICE CELL PARAMETERS

by

R.E. KAY

Chalk River Nuclear Laboratories

Chalk River, Ontario

June 1976

ATOMIC ENERGY OF CANADA LIMITED

LATTICE MEASUREMENTS WITH 37-ELEMENT BRUCE REACTOR

FUEL IN HEAVY WATER MODERATOR:

DETAILED LATTICE CELL PARAMETERS

by

R.E. Kay

R e a c t o r Phys i cs B ranchChalk R i v e r N u c l e a r L a b o r a t o r i e s

Cha l k R i v e r , O n t a r i oJune 1976

AECL-5307

Mesures de réseau avec le combustible de 37 éléments destiné

aux réacteurs de Bruce où le modérateur est de l'eau lourde:

Paramètres détai l lés des cellules de réacteur

par

R . E . Kay

Résumé

De1" expériences ont été effectuées dans l'ensemble cr i t ique ZED-2pour"déterminer les paramètres de réseaux typiques en diverses positionsdans la cel lu le centrale des réseajx du combustible de 37 éléments duréacteur Bruce où le modérateur est de l'eau lourde. Les mesures ontété effectuées dans le pas de réseau carré de 28.58 cm du réacteur Bruceen u t i l i san t comme caloporteurs de l ' a i r et de l 'eau lourde.

Le présent rapport décri t ces expériences et i l présente les résultatsen fonction des:

répart i t ions détaillées de l ' a c t i v i t é relat ive du cuivre,

rapports re la t i f s d 'ac t i v i té de Tindium-manganèse et du lutétium-manganèse,

paramètres de Westcott r et T , etn

rapports de conversion i n i t i a l e et des rapports de f ission rapide.

Les valeurs expérimentales sont comparses aux résultats des calculsDSM et PU WIMS.

L'Energie Atomique du Canada, LimitéeLaboratoires Nucléaires de Chalk River

Chalk River, Ontario

Juin 1976AËCL-5307

LATTICE MEASUREMENTS WITH 37-ELEMENT BRUCE REACTOR FUEL IN

HEAVY WATER MODERATOR: DETAILED LATTICE CELL PARAMETERS

by

R.E. Ka"

ABSTRACT

Experiments have been performed in the ZED-2 criticalfacility to determine typical lattice parameters atvarious positions in the central cell of lattices of37-element Bruce reactor fuel in heavy water moderator.The measurements were made at the Bruce reactor squarelattice pitch of 28.58 cm using air ;nd heavy water as"coolants".

This report describes these experiments and presentsthe results in terms of:

detailed relative copper activity distributions,indium-manganese and 1utetium-manganese relativeacti vi ty rati os ,the Westcott parameters r and T n , andinitial conversion ratios and fast fission ratios,

Experimental values are compared with the results ofDSN and P1J WIMS calculations.

Reactor Physics BrancnChalk River Nuclear Laboratories

Chalk River, Ontario

June 1976

AECL-5307

TABLE OF CONTENTS

Page

1. INTRODUCTION 1

2. DESCRIPTION OF FUEL, LATTICES AND COOLANTS 2

2.1 Lattice Configurations 22.2 Fuel Assemblies 2

2.3 Demountable Fuel Bundle 3

3. EXPERIMENTS . 5

3.1 Copper Activity Distribution Measurements 53.2 Lutetiurn-Manganese, Indium-Manganese

Activity Measurements 113.3 Relative Conversion Ratio and Fast Fission

Ratio Measurements 164. WIMS EXPERIMENTAL COMPARISONS 26

4.1 General 264 . 2 F a s t F i s s i o n R a t i o , 6 264 . 3 R e l a t i v e C o n v e r s i o n R a t i o , C 274 . 4 R e l a t i v e U 2 35 F i s s i o n Rate D i s t r i b u t i o n 284 . 5 R e l a t i v e Copper A c t i v i t y D i s t r i b u t i o n 284 . 6 L u t e t i u r n - M a n g a n e s e R a t i o s 29

5. SUMMARY 30

6. ACKNOWLEDGEMENTS 31

7. REFERENCES 32

LIST OF TABLES

Table 1 Lattice Data

Table 2 Summary of Normalized Copper Activities

Table 3 Detailed Normalized Copper Activities

Table 4 Foil Relative Activities and Ratios

Table 5 Westcott Spectrum Parameters

Table 6 Values of 5, C, and Relative U 2 3 5 Fission Rates

Table 7 Typical WIMS Specifications

Table 8 WIMS/Experimental Ratios

Table 9 WIMS/Experimental Ratios

LIST OF FIGURES

Fig. 1: Typical Lattice Arrangement in ZED-2

Fig. 2: Cross Section Through 19-Element U Metal Fuel Assembly

Fig. 3: Cross Section Through 37-Element Fuel Assembly

Fig. 4: 37-Element Bruce Fuel Bundle

Fig. 5: Demountable Bundle and Components

Fig. 6: Demountable Package and Components

Fig. 7: Showing Detector Locations

Fig. 8: Sector Foil Cutting Scheme

Fig. 9: Fuel Region Average Copper Activities; Heavy WaterCool ant

Fig. 10: Fuel Region Average Copper Activities; Air Coolant

Fig. 11: Copper Activity Distributions; Heavy Water Coolant

Fig. 12: Copper Activity Distributions; Air Coolant

Fig. 13: Copper Activity Around a Fuel P^n: Heavy Water Coolant

Fi -]. 14: Copper Activity Around a Fuel Pin: Air Coolant

Fig. 15: Arrangement of Uranium Foils

Fig. 16: P(t) Versus Time

LATTICE MEASUREMENTS WITH 37-ELEMENT BRUCE REACTORFUEL IN HEAVY WATER MODERATOR:

DETAILED LATTICE CELL PARAMETERS

by

R.E. Kay

1 . INTRODUCTION

This report describes experiments performed in thecritical facility ZED-2 to determine typical latticeparameters in the central cell of a latt.ce of 37-eiementBruce reactor fuel in heavy water moderator. Measurementsware made at the Bruce reactor square lattice pitch of11.25 inches (28.58 cm) using air and heavy water as"cool ants".

The parameters measured were:detailed relative copper activity distributionsthroughout a lattice cell,indium-manganese and 1utetium-manganese activityratios relative to a thermal reference location,and hence the Westcott spectral indices T n andr/VVinitial conversion ratios and fast fission ratiosthroughout a fuel bundle.This report is complemented by Reference 1 which

describes buckling and related measurements onlattices of Bruce reactor fuel in ZED-2.

Section 4 presents a comparison between the resultsof DSN and P U WIMS calculations and comparable measureddata.

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2. DESCRIPTION FOR FUEL, LATTICES AND COOLANTS

2.1 Lattice Configurations

Figure 1 is typical of the lattices used. Usually

they were composed of a central region of 69 assemblies

each containing five 37-element Bruce reactor fuel bundles,

on loan from Ontario Hydro, with either air or heavy

water "coolant"; surrounded by a ring of 28 assemblies

containing 19-element natural uranium metal fuelv ' with

heavy water coolant. This outer ring of fuel was required

to increase system reactivity.

The assembly at the centre of the lattice contained

a special demountable simulation of a Bruce fuel bundle

with two regular Bruce bundles both above and below. All

the detailed activation measurements were made within or

about this demountable bundle.

All fuel assemblies were arranged in a square closed

centered lattice at the Bruce reactor pitch of 28.58 cm

(11.25 inches). For several experiments six fuel assemblies

in the southwest corner of the lattice were removed to

produce a "thermal pit" wherein reference foils associated

with various neutron spectrum parameter measurements could

be irradiated at least 40 cm away from the nearest fuel

assembly.

Details of the actual core configurations, moderator

conditions and coolant used in each experiment are presented

i n Table 1 .

2.2 Fuel Assemblies

A section through one of the twenty-eight 19-element

uranium metal fuelled assemblies used to increase the

reactivity of the lattice is presented in Fig. 2. See

-3-

Reference 2 for details.

Sixty-eight of tho 69 simulated Bruce fuel assemblies

consisted of five 37-eiement Bruce reactor fuel bundles

stacked within concentric aluminum pressure and calandria

tubes. Figure 3 is a cross section through such an

assembly while Figure 4 illustrates a Bruce fuel b u n d l e .

Each fuel element contains a stack of natural uranium

oxide fuel pellets (12.1 mm diameter, density 10.50 g/cm 3)

within a Zircaloy-4 sheath of wall thickness 0.4 mm and

I.D. 12.2 mm. Fuel stack length was about 48 cm; overall

bundle length 49.5 cm.

Eighteen, twelve, six and one fuel elements were

assembled on circles of diameters 8.661 cm, 5.751 c m ,

2.977 cm and 0 cm respectively. A bundle was formed by

welding these 37-elements to two Zircaloy-4 end plate

"spiders"; see Figure 4. Being production fuel these

bundles incorporated such features as dished pellet e n d s ,

fission product gas plenum space, bearing pads etc.

Dimensions of the simulated Bruce pressure-calandria

tube assembly are given in Figure 3. Although both tubes

and end plate were of 1050 aluminum alloy, s u ^ e q u e n t

reactivity measurements on these assemblies^ ' gave higher

than expected neutron absorption properties.

2.3 Demountable Fuel Bundle

The fuel assembly at the centre of the lattice dif-

fered from the other 68 simulated Bruce assemblies only

because the middle (third) Bruce fuel bundle was replaced

by a special demountable simulated Bruce bundle. All

detailed activation measurements were made within or about

this bundle.

Figure 5 illustrates this demountable bundle in a

partially assembled s t a t e , and the various components of a

demountable fuel pin. Each fuel sheath was nominally of

the same material, diameter and wall thickness as that used

in the Bruce fuel. Thirty of the thirty seven fuel

elements contained UO2 pellets to the same material and

dimensional specifications as used in Bruce reactor fuel

(including dished pellet ends etc . ) . These thirty

elements were fitted with welded end caps which incorporated

a screw threaded projection.

The remaining seven elements (A to D3 of Figure 7) could

be completely disassembled to enable foils to be positioned

between fuel pellets. The associated aluminum end caps with

0-ring seal were a watertight push fit into the Zr-4 sheaths.

Although the fuel used in these seven elements was UO2 to

the Bruce fuel specifications, all pellets were flat ended

and of reduced diameter (11.73 m m ) . These reduced diameter

pellets could be packaged, together with foils, within very

thin walled (0.13 mm) 2S aluminum cans of I.D. 11.8 mm,

length about 10 cm. The resulting package illustrated in

Figure 6 was convenient to handle, obviated foil or pellet

jamming problems, and allowed accurate alignment of the

foils with adjacent fuel pellets.

The demountable bundle was assembled from the thirty

welded elements plus the seven demountable elements by

bolting a number of the threaded end caps through aluminum

end plates. These end plates ensured that the positions

of the thirty-seven fuel elements were as in a Bruce fuel

bundle.

-5-

3. EXPERIMENTS

For convenience the various measurements are discussedin three completely separate sections dealing respectivelywith: detailed copper activity distributions throughout alattice cell, detailed In-Mn and Lu-Mn relative activityratio measurements through a cell, and initial conversionratio and fast fission ratio measurements within a fuelbund!e .

3.1 Copper Activity Distribution Measurements

3.1.1 General Description

Figure 1 is typical of the lattices used in this seriesof experiments. The six fuel assemblies associated withthe thermal pit were always present; their removal wouldhave produced unacceptable perturbation of the macroscopicthermal neutron distribution across the core.

In these experiments measurements of detailed copperactivity distributions throughout the various regions ofthe central lattice cell were made using copper foils andwires (see Figure 7 ) . All copper detectors were locatedat about the elevation of the mid height of the demountablefuel bundle well away from bundle ends. Although groups ofdetectors were at slightly different elevations they wereall located within ~30 cm of the peak of the axial cosinethermal flux distribution, and relative height correctionfactors were modest.

Circular copper foils 11.7 mm (0.462 in.) diameter and0.14 mm thick were inserted between UO2 fuel pellets inelements A to D3 and on the outside of the calandria tubein four locations; see Figure 7. To prevent contamination,during irradiation the foils located between fuel pelletswere canned in 0.03 mm thick aluminum foil.

-6-

As illustrated in Figure 7 a sector-shaped copper

foil of thickness 0.13 mm occupied a representative 1/12

of the coolant flow area. After irradiation this sector

was cut into a number of similar size pieces (as in Fig. 8)

and counted together with the various circular foils and

wi res .

Long (-40 mm) copper strips, ~5 mm wide and 0.13 mm

thick were taped around the outside of the sheaths of

several fuel elements. After irradiation each strip was

cut into eleven pieces and counted together with the other

copper detectors.

Detailed activity distributions in the moderator

region of the cell were made using 10 mm long, 0.76 mm

diameter copper wires taped to the three arms of a light

aluminum framework which extended out into the moderator

of the central cell; see Figures 1 and 7. This framework

had arms aligned east, northeast and north-south, which

lay respectively along a line joining two nearest neighbour

fuel assemblies, along a diagonal of the lattice cell, and

along a cell boundary.

Height correction factors were evaluated from a cosine

least squares fit to the data obtained from the activation

of a series of 11.3 mm (0.444 in.) diameter, 0.25 mm thick

circular copper foils spaced at 1.0 cm intervals, suspended

within an aluminum thimble located at cell edge position Q

of Fi gure 1 .

Note that in each experiment the alignment of fuel

elements A to D3, the strips around a fuel element, the

sector foil, the four foils on the calandria tube, and the

aluminum framework, relative to the reactor datum direction

north (see Figures 1 and 7) was Maintained.

-7-

3.1.2 Determination of Copper Foil and Wire Activities

The 12.9 h Cu 6^ gamma ray activity of all the variouscopper detectors was determined using an automatically re-stacking sample changer-counter system utilizing two 51 mmdiameter by 25 mm thick Nal (la) scintillation detectors.To ensure accurate location in the counter and reproducibi1ityof count rate, the various detectors were held in lucitepallets.

The output from the two detectors was recorded on atypewritten sheet, and on paper tape which was later fed toa CDC-6600 computer for data reduction. The computer pro-cessing made numerous checks on count reproducibi1ity,corrected for counter dead time, corrected all results toa common zero time etc., and produced relative specificshape dependent activities A , (and statistical andreproducible counting errors) for a position x in the cell.Sufficient counts were accumulated for each detector togive a statistical accuracy of $0.5%.

3.1.3 Foil and Wire ntercalibrations and Corrections

The various copper detectors had different shapesand hence different neutron flux depression effects, dif-ferent counter sensitivities, etc. Consequently the measuredrelative specific copper activities A v, for the differentdetectors are not exactly consistent. Therefore in separateexperiments several samples of each shape of detector wereirradiated on a rotating wheel located in a "reference"thermal pit region within ZED-2 (see Figure 1 ) , then countedand analyzed as in Section 3.1.2 to produce relative specificactivities A Ri- The previously determined shape dependentspecific activities A , were normalized to the correspondingshape reference activity A R, to produce a consistent set of

-8-

shape corrected specific copper activity data A v , i.e.A

Ax = Ax' / AR"*As mentioned in Section 3.1.1 the various copper

detectors were at different elevations close to thedemountable bundle mid length. Relative height correctionfactors were evaluated from a cosine fit to the copperdistribution of thimble Q; such height corrections were

-9-

The average activity of an individual ring of elements

(A c, A D ) is defined in a similar manner.

A is the average activity of the strip wrapped

around a fuel sheath relative to the average activity of

the fuel region of that element.

The moderator average copper activity A was obtained

by a numerical integration method using the copper data

in the three directions through the moderator.

Based upon counting statistics the error associated

with these normalized activities is < 0 . 4 % for all foils

between fuel pellets, on the calandria tube and at the

cell edge, ~~\% for the coolant average, and -0.6% at the

fuel element sheath. An error in normalized moderator

average activity - 0 . 8 % was obtained.

Looking in detail at the results of each experiment

we observe the very good agreement between activities

measured at nominally identical or very similar cell

locations. For example, the activities of fuel elements

Cl and C 2 , and of Dl, D2 and D3 are in close agreement, as

are the values for the four locations around the calandria

tube, and at the two diametrically opposite cell edge

locations (viz foil in thimble Q and wire on the east

framework). No significant discrepancies or systematic

trends are apparent.

Figures 9 to 12 illustrate bundle average activities

and detailed activities in the moderator region. As expected,

thermal flux gradients are steeper through the heavy water

cooled bundle than the air cooled bundle.

Figures 13 and 14 are polar plots of normalized

activities around the sheath of representative fuel pins.

The circumferential activities clearly illustrate the

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macroscopic thermal flux depression radially through

the bundle. The local perturbing effects cf individual

adjacent fuel elements (i.e. scalloping) is not clearly

discernible within the accuracy of the measurements

(-0.6%) and the limited angular resolution of -30°.

Some of the results indicate slight angular misalignment

of the copper foils.

-11-

.1. 2 Luteti urn-Manganese , Indium-Manganese Activity Measurements

3.2. Method

Integral information on neutron spectrum shape in

the thermal and epithermai energy ranges can be obtained

from Lu-Mn and In-Mn relative activities. This is pos-

sible because of the significantly different thermal and

epithermal cross-section behaviour of these materials:

Lu^76 has a very non-l/v cross-section behaviour in the

thermal energy range, exhibiting a broad absorption

resonance at 0.142 eV; I n 1 1 5 which is a nearly 1/v

absorber at low thermal energies, has a large absorption

resonance at 1.46 eV; Mn55 is essentially a 1/v absorber

up to energies -few eV.

Experimentally, thin In-Al and Lu-Mn-Al alloy foils

were irrad'ated together at various cell locations x and

on a rotating wheel located in the thermal pit reference

position wh?re the neutron spectrum was essentially a

Maxwellian distribution at the physical temperature of the

heavy water moderator. Full details of the experimental

method and method of analysis are presented in Referenceno .

The I n 1 1 6 , M n 5 6 and L u 1 7 7 activities of the foils

were determined and expressed in terms of both relative

reaction rate ratios R, a parameter useful for comparison

with cell calculational schemes such as HAMMER and WIMS,

and the Westcott indices r and T n as used in lattice

recipe codes. These parameters were determined using

the following expressions.

-12-

Gr 5o) Mn

£ r / V V P Grso) M

In

Mn

1)

i [A, /AM ],Lu L Lu' MnJx Mn 2)

where A is the specific activity, x refers to the positionin the lattice cell under study, and P refers to the thermalflux reference position; g for Lu and for In are knownfunctions of the neutron temperature T .

Note that in these expressions allowance is made forthe fact that r/T /T may be non-zero at the referenceposition. Its value there was determined by measuring theCd ratio for thin In-Al foils; (rv is so smal1 thatany uncertainty in this measurement introduces a negligibleerror into the determination of (r/T / T _ ) x - Values of G rs Qfor the various detectors were obtained from Cd ratio measure-ments and were based on the value s = 18.8 for In. G^values were calculated using Hanna's method for thermal self-shielding^ .

3.2.2 Measurements

Figure 1 represents the type of lattice used in these

experiments. Note that the six fuel assemblies in the

southwest corner of the lattice were removed to form a

-13-

"thermai pit" reference spectrum position.

The activation measurements were made using 0.13 mm

thick, 11.7 mm diameter 1% In-Al alloy and 10% Lu-5% Mn-AI

alloy foils. Pairs of these foils canned in 0.03 mm thick

aluminum foil (to prevent contamination), were inserted

between fuel pellets in elements A to D3 (see Figures 6 and 7)

of the demountable bundle, on the north and south faces of

the caiandria tube, in thimble T, and on a rotating wheel

located in the "thermal pit". All foils were located

approximately at the elevation of the mid point of the

demountable bundle.

The Cd ratio for In was measured in the thermal pit

using 11.3 mm diameter, 0.13 mm thick In-Al foils and Cd

and Al boxes with 0.76 mm thick walls.

Relative height corrections were made using the data

obtained from copper foils held within an aluminum thimble

at location Q, Figure 1.

3.2.3 Determination of Foil Activities

All foils were counted using the equipment described

in Section 3.1.2, the data being processed by the CDC-6600

computer to produce relative specific activities.

The 54 minute I n ^ ^ activity was determined by counting

foils for 2 - 3 hours beginning ~1 hour after reactor shut-

down with the counter bias set at -40 keV. The 2.58h M n 5 6

activity of the Lu-Mn-Al foils was determined by counting

for several hours (commencing ~2 hours after reactor shut-

down) with a counter bias of 500 keV to exclude all Lu

acti vi ty.

Two days after irradiation the Lul?7 activity was

determined using a counter bias of -40 keV. The waiting

period ensured the full decay of both the M n 5 6 and the

-14-

3.7 h L u ^ 6 -jsomer activities. A correction was made

for the 2.2 x 1 0 1 0 a L u 1 7 6 activity by counting the foil?

before irradiation.

3.2.4 Foil Intercalibration

The alloy foils used were not uniform so that therelative concentrations of In, Lu and Mn were determinedby irradiating them on a rotating wheel in a thermalcolumn of the high flux reactor NRU, then counting themas described above to determine the relative activitiesof the relevant isotopes.

3.2.5 Experimental Results and Discussion

Table 4 lists foil specific activity ratios relativeto the corresponding ratio at the reference position(i.e. RJ," and Rhu)> a n d individual foil relative activities.Although these activity ratios are readily comparable withcalculable quantities, their significance can be appreciatedmore readily by considering the corresponding Westcottindices listed in Table 5.

The statistical counting error of the individual foilactivities was typically -0.15%. Because the ratios RM

Luand RM contain eight foil activities (including relativesensitivity measurements) the resulting errors in theseratios are ~j^ 0.4%. These errors lead to errors in r/T n/T Qand AT of ~+_ ]% and -+_ 2 ° c respectively.

Although the spectrum parameters of Table 5 and therelative activity ratios of Table 4 show good consistencybetween nominally identical cell locations (e.g. fuelelements Dl and D 3 ) , systematic effects due to the pertur-bation of the macroscopic thermal flux distribution across

-15-

the core by the thermal pit can be detected in theresults of Table 4. Note that the individual foil relativeactivities of Table 4 have been corrected to a constantelevation; the relative height corrections were small (

-16-

3. 3 Relative Conversion Ratio and Fast Fission Ratio Measurements

3.3.1 Method

In a CANDU*type reactor a few percent of all fissions take

place in U 2 3^. The calculable and measurable parameter

related to these fast fissions in U 2 3 8 is the fast fission

ratio g.

5 = [ U 2 3 8 fissions/U 2 3 5 fissions]x

The method used to determine g is described in detail

by Bigham in Reference 5. In this report only a bare

outline of the method is presented.

Thin foils of natural U and depleted U were irradiated

between fuel pellets within the demountable bundle, and the

induced fission product -y-ray activities were determined.

Using the notation of Bigham ' 5 is related to the

gamma ray measurements through the expression,

& = P(t) RT(t) ...3)

where P(t) is a time dependent calibration factor obtained

by irradiating the same foils in a reference neutron flux

where 6 is known from direct fission chamber measurements,

and Ry(t) is a time dependent ratio related to the measured

foil activities as follows:

D -...4)

f 8 N - D

where D = specific activity of the depleted U foil and

N = specific activity of the natural U foil.

* Canada Deuterium Uranium

-17-

,t5)The factors f$ and f8 are as defined by Bighan

+s = T: and fR =

(nQe) 5

where n is the number of atoms per unit mass of uranium,Q is a flux perturbation factor that depends on

the geometrical arrangement of the foils inthe test fuel assembly (see pages 26 - 28 ofreference 5 ) , and

E is the counter efficiency for fission product

Y-rays and includes Y" r ay s self-absorption.

The subscripts 5 and 8 refer to U235 and U238 and super-

scripts u and d to natural and depleted uranium respectively.

The long-term reactivity behaviour of a naturaluranium fuelled reactor is very significantly affected bythe production of fissile Pu 2 3^ formed from neutron capturein U238.

N p239 ^ _ _ ^ pij239^ > N p1 1/2 = 23.5 min ' 1/2

It is therefore important to be able to calculatereliably some parameter related to U^38 captures. Asuitable parameter which is calculable and can be measured,is the relative conversion ratio C where

C = [U 2 3 8 captureb/U23^ fjssions]v[U238 captures/U235 fissions]p

and x and P refer to a fuel location and a thermal referencelocation respectively.

-18-

The experimental technique used to determine C hasbeen described in detail by Tunnicliffe et al in Reference6, and only a brief outline is presented in this report.

Experimentallyi natural uranium metal foils wereirradiated between fuel pellets in the demountable bundle,and on a rotating wheel in the thermal pit referenceposition. The counting technique employed to determinerelative U 2 3^ captures (i.e. Pu 3^ productions) was tomeasure relative N p 2 3 9 decays by the coincidence countingof 106 keV Y-rays and fluorescent X-rays. This coincidencemethod suppresses the counts from interfering fission pro-duct and natural background activities. Relative U 2 3 5

fission product y activities were determined as in the fastfission ratio measurements.

Therefore in terms of the measured quantities therelative conversion ratio

r _ [Np239 coincidence activity]x / [Np2 3 9 coincidence activity]p

[ U t 3 5 fission product act.] x / [U2 3 5 fission product act.] p

3.3.2 Measurements

Note, that these measurements were made at the sametime as the spectrum parameter measurements of Section 3.2.2",therefore the lattice used was that of Fig. 1, with sixassemblies removed to form a "thermal pit" referencespectrum location, and with a set of copper foils at 10 cmintervals within thimble Q.

The activation measurements were made using 0.07 mm thick,natural U metal and 0.12 mm thick depleted uranium metalfoils, both 11.7 mm in diameter. The U 2 35 content of thedepleted U was less than natural by a nominal factor of 20.In each of fuel pins A to D3, foils were arranged in therelative positions illustrated in Fig. 15; note that thenatural U foil was always located opposite the four holes

-19-

in the aluminum can (Fig. 6 ) . Each depleted foil was

wrapped (to prevent contamination and oxidation problems)

and separated from adjacent UO2 fuel pellets by 0.03 mm thick

aluminum "catcher" foils. These thin catcher foils shielded

the depleted foil from fission fragments originating in the

adjacent fuel pellets; they were removed before counting.

The bare natural U metal foil was separated from

adjacent pellets by two pairs of catcher foils. Before

counting,the outer catcher foils (contaminated with fission

fragments originating in adjacent UO2 pellets) were discarded;

the inner catcher foils bearing fission fragments originating

in the U metal foil were retained.

Because the natural U foil was associated with measuring

11^38 capture rates it was important to align this foil with

the adjacent U 0 2 pellets. This alignment was achieved by

eye, observing relative foil and pellet circumferential

edges through the four holes in the aluminum can.

Similar natural foils were irradiated on

a rotating wheel in the "thermal pit" reference spectrum

location. Also on this rotating wheel were Al and Cd covered

natural uranium foils used to determine the Cd ratio for

l|238 captures' , and two 11.5 mm diameter, 0.25 mm thick

fully enriched (93% U 2 3 5 ) 5% U-Al alloy foils used to make

the correction for spurious coincidences arising from fissiort

product (F.P.) - y - r a y s ^ .

Because of the time dependent properties of Ry(t) and

P(t) the ZED-2 irradiation was for a specific time (60 min)

and the time of shutdown was carefully noted.

3.3.3 Determination of Foil Activities

Because the counting equipment used is described in

detail in References 6 and 8, only a brief description will

be given here.

-20-

All foils were counted in an automatically restackingsystem with two Nal (T£) crystals mounted on photomultipliertubes. For fission product (F.P.) y-ray counting, integraldiscrimination at 1.25 MeV was used.

For N p " 9 counting, the two photomul tipl ier outputswere fed to independent linear amplifiers and single channelpulse height analyzers set to cover a peak at ~100 keV(width -30 keV) which contained roughly equal numbers ofunresolved 106 keV y-rays and Pu^39 x-rays with energiesfrom 99 to 118 keV. The latter arise from the internalconversion of y-rays in the Np239 decay. Coincidencesbetween the 106 keV y-rays and the X-rays were determinedby feeding the outputs from the single channel analyzersinto a coincidence circuit with a resolving time of 0.5 ys.

The foils were counted in turn starting from the firstto the last and then in the reverse order. This constitutedone counting cycle in which each sample was counted twice.This method of counting obviated the need for decay cor-rections. Individual count times were 40 s for fissionproduct counting* 200 s for coincidence counting.

3.3.4 Data Analysis

The counting results, recorded on paper tape, were pro-cessed by the CDC-6600 computer using the program ICRAFF toprovide time dependent relative specific F.P. counts andcoincidence counts, corrected for background, dead time,and spurious coincidences. See Ref. 7 for additional featuresof counting and data processing.

Typically the sequence of counting and analysis eventsprogressed as follows .

F.P. and coincidence background count all foils to beused for ~24 hours. Data analyse to produce individual

foil background counts per second.

-21-

Perform ZED-2 irradiation for a specific time (60 min)

and note reactor shutdown time.About ~3h after reactor shutdown begin F.P. counting ofnatural and depleted foils. Count for ~20 hours. Data

analyse by ICRAFF.About 2 days after shutdown (by which time F.P.relative activity has significantly decreased) begincoincidence counting of natural uranium foils (and

enriched foils). Also record F.P. counts to enablethe spurious coincidence correction to be evaluated^ .Count for -72 hours. Data analyse by ICRAFF.

The resulting time-dependent F.P. and coincidencespecific activities were then processed by a CDC-6600 pro-gram FICAL to produce the desired parameters 5 and C. Fordetails rf the stages in this analysis see Ref. 7. A briefstep-by-step explanation is presented below.

The F.P. counting of both the natural and depletedfoils included a small but not insignificant component(typically

-22-

Q, were applied.

The gap correction factor Q = Q5/Qo> was evaluated

using expression 37 and the data of reference 5.Since Q is a function of 6 it was evaluated withinFICAL by an iterative process. Typically Q ~ 1.013to 1.023.

The factor fg was evaluated using the techniques anddata of Reference 5, giving (for a typical foil)

n d Q d e d

fn = -r " -r • -4 = 1-0068 x 1.0000 x 0.9745 = 0.9811O U n U U

n8 Q8 C8

The factor f5 was evaluated^ ' by calculating e^/e^,assuming Q5/Q5 = 1.000, and obtaining njj/nE; by F.P.counting^ ' of several natural and depleted foilsirradiated in a thermal reference location.For typical foils

d nd d

fc = - 7 — • -f = 0-05805 x 1.000 x 0.9745 = 0.05657.n5 Q5 e5

The time dependent factor P(t) was available from(9)previous measurements made by 0kazakiv ' using

identical foil material of the same thickness, inthe form of circular foils 13.0 mm diameter; see

Figure 16.For each foil location A to D3, a value of R and hence6 was evaluated for each counting cycle, using expressions3) and 4 ) . At each location (including fuel and ringaverages) a single weighted mean value of 6 (and associatedstandard deviation) was then evaluated. These values are

1i sted i n Table 6.

At each counting cycle the relative total F.P. activityof each natural foil (Al corrected) was corrected forU238 F.P. activity by multiplying by 1/O+R h to giveU 2 3 5 F.P. activity relative to the thermal referenceposition. In applying this correction it is assumed

-23-

2 38that U F.P. activation at the thermal pit location

was negligible: this is reasonable since the measured

F.P. Cd ratio for U natural foils at this location was

>3O00.

At each location (including fuel and ring averages) a235single weighted mean value of relative U activity

(and associated standard deviation) was then evaluated.

The natural uranium coincidence activity (corrected

for spurious coincidences) of each foil was corrected

for differential 100 keV Y-ray self absorption effects.

Such effects are small, a typical overall variation

in individual foil thickness (mass) of -10% resulting

in a

-24-

3.3.5 Experimental Results and Discussion

Table 6 summarises the results of these experiments.

As might be expected values of 6 are higher and values of

C are lower, for the air cooled cell than for the heavy

water cooled cell, and the U 2 3 5 fission distribution through

the bundle is flatter in the air cooled cell.

Fuel bundle and ring average values of (D-f 5N) and

(fgN-D), and hence 6, were evaluated using the number of

fuel rods of a given type as the weighting factor.

The standard deviation of the weighted mean values of

6 were typically

- 2 5 -

Fue l bund le and r i n g average v a l u e s o f C were o b t a i n e d

u s i n g b u n d l e Cor r i n g ) average v a l u e s o f r e l a t i v e c o i n c i -

dence and f i s s i o n p r o d u c t a c t i v i t i e s , where r e l a t i v e

numbers c f f u e l rods o f a g i v e n t y p e were used as w e i g h t i n g

f a c t o r s .

The s t a n d a r d d e v i a t i o n o f t h e w e i g h t e d mean v a l u e s o f

C was

-26-

4. WIMS EXPERIMENTAL COMPARISONS

4. l General

WIMS calculations were performed using both the

DSN and PIJ options, to obtain values comparable with the

measured parameters of Tables 2, 4 and 6.

These calculations were made using the CRNL version

of WIMS-D2 as of January 1976, with a 69 energy group, 80

nuclide, library tape designated "Winfrith Reel # AEC 331".

The actual cell geometric specification and the isotope

identifiers used are presented in Table 7.

Calculations were performed for the heavy water purity

and temperatures of the corresponding experiments.

Note that the basic experimental measurements are of

real foil relative reaction rates obtained from fr Is

located between slightly undersi ze fuel pellets in 7 elements

of a bundle having normal size fuel in its remaining 30

fuel elements (see Section 2.3 for details). Some of the

analyzed experimental data of Tables 2 to 6 has been cor-

rected for local perturbation effects (viz

-27-

6 W = [U2 3 8fissions/U 2 3 5fissions] x

Table 8 lists DSN and P1J WIMS calculated values of

-28-

235 238The N and a are U and U number densities and

Maxwellian averaged (capture or fission) cross sections

respectively. The cr were evaluated by a computer pro-

gram MAXWIM2, using the 69 energy group cross sections

of the WIMS library tape: -

V c V f = 4-7625 x 10~3 a t 2 9 5 ° A -

Table 8 lists DSN and PIJ WIMS calculated values of

C divided by the corresponding measured quantity of Table

6. There is good agreement between calculated values and

experiment. The -0.7% underprediction in bundle average

C is equivalent to ~+3 mk in reactivity.

2354.4 Relative II Fission Rate Distribution

The WIMS output allows a direct evaluation of the235ratio of U fissions at a location x, to fuel average

Uc D fissions. Table 8 lists WIMS calculated values of

normalized U fission rate, divided by the corresponding

measured quantity of Table 6.

Within the fuel bundle these data (and the corresponding

C u 6 3 data of Table 9) illustrate that the DSN calculations

predict too severe a (thermal) flux depression through the

bundle: the corresponding PIJ data are in excellent agree-

ment with experiment.

4.5 Relative Copper Activity Distribution

Table 9 lists DSN and P1J WIMS calculated values of

normalized Cu absorption rates, divided by the corresponding

measured quantity of Table 2. These data illustrate that

the DSN calculations predict too severe a (thermal) flux

depression throughout the whole cell, while the corresponding

P1J calculated data are in very good agreement with

-29-

experiment.235Within the fuel bundle the copper data and the U

fission rate data of Table 8 are very comparable.

4.6 Lutetiurn-Manganese Ratios

The WIMS calculated 1utetium-manganese ratio is:

I MnJ W [a i /o M 1L m.Lu' m,MnJ

where A are activation rates and the a are Maxweiiianaveraged cross sections at the physical temperature ofthe moderator of the experimental "thermal pit".

V L u ' V M n = 2 6 5 ' 4 a t 2 9 5 ° A

Table 9 lists WIMS calculated R ^ divided by the cor-responding data of Table 4. WIMS-Experimental agreementis very satisfactory. A 0.5% overprediction of fuelaverage R., corresponds to an underestimate in reactivityof only -0.3 mk.

-30-

5. SUMMARY

This report describes experiments performed to determine

typical cell parameters in lattices of 37-element UO- Bruce

reactor fuel. Measurements were made at a single lattice

pitch, namely 11.25 inches square (28.58 c m ) , using heavy

water and air as coolants.

Measurements of relative copper activities were made at

many locations within a fuel bundle and at other locations

in a eel 1.

Integral information on neutron spectrum shape in the

thermal and epithermal energy ranges was obtained from

measurements of relative Lu-Hn and In-Mn activities at

various cell locations. The measurements were expressed in

terms of both foil activity ratios relative to thermal pit

activities (R) and the conventional Westcott indices rv/Tn/TQand T n .

Measurements of the fast fission ratio 6 and relative

conversion ratio C were made within seven elements of the

fuel bundle, using t!.e techniques and methods developed by

Tunnicliffe, Bigham and others. Brief descriptions of these

techniques and the methods of analysis are presented.

WIMS calculations were performed using both the DSN ani

PIJ options. Calculated values of 6, C, Lu-Mn relative235ratios, and copper and U fission distributions are

compared with corresponding experimental data.

-31-

6. ACKNOWLEDGEMENTS

The fuel bundles used in these experiments were loaned

by Ontario Hydro, through the good offices of R.V. Belluz

(Power Projects).

The author wishes to thank the many persons involved

in performing and analyzing the experiments and subsequently

producing this report. In particular to P.D.J. Ferrigan,

E.J. Pleau and D.J. Roberts at the ZED-2 reactor, D. Kettner

and F. Mazzone who assembled the demountable bundle and

supervised the counting, D. Wright who produced the figures

for this document, and G.D. Clark who typed the manuscript.

-32-

7. REFERENCES

1. A. Okazaki, R.T. Jones; "Buckling Measurements ofBruce 37-Element UO2 Fuel Bundles in ZED-2",Atomic Energy of Canada Limited, UnpublishedInternal Report, CRNL-1450 (1976).

2. K.J. Serdula, R.E. Green, "Lattice Measurementswith 19-Element Natural Uranium Metal Assemblies",Atomic Energy of Canada Limited, Report AECL-2516(1965).

3. C.B. Bigham, et al ; "Experimental Effective FissionCross-Sections and Neutron Spectra in a UraniumFuel Rod", Atomic Energy of Canada Limited, ReportAECL-1350 (1961).

4. G.C. Hanna, "The Neutron Flux Perturbation Due to anAbsorbing Foil", Nucl. Sci. Eng. 15, 325, (1963).

5. C.B. Bigham, "Measurements of Fast Fission Ratios inNatural Uranium", Atomic Energy of Canada Limited,Report AECL-2285 (1965).

6. P.R. Tunnicliffe et al , "A Method for the AccurateDetermination of Relative Initial Conversion Ratios",Nucl. Sci. Eng. 15, 268, (1963).

7. P.W. deLange et al , "Experimental Initial Conversionand Fast Fission Ratios for Clusters of Natural Uand UO2 in D2O" , Atomic Energy of Canada Limited,Report AECL-2636 (1966).

8. B.G. Chidley et al , "Initial Conversion Ratios andFast Fission Factors in Heavy Water Natural UraniumLattices Using 19-Element UO2 Fuel Rods", Nucl. Sci.Eng. 17, 47 (1953).

9. A. Okazaki, Private communications.

10. O.R. Askew et al, "A General Description of the LatticeCode WIMS", Journal of B.N.E.S., pp. 564 - 585,October (1966).

rABLE

LATTICE DATA: PITCH 28.58 cm SQUARE

EXPERIMENT

Copper

Activation

Lu/Mn

In/Mn«5,C

COOLANT

/ Ai r

\ HeavyI Water

/ Ai r

< Heavy( Water

DATE

21/4/75

9/4/75

25/3/75

18/3/75

TOTAL NO.FUEL ASSEMBL

97

97

91

91

OFIES TEMP.

21 .51

21 .50

21 .75

21 .83

MODERATOR

°C PURITY ATOM 5

99.711

99.713

99.711

99.712

BUCKLING

>. a V 2

1 .337

1 .192

1 .303

1 .153

DATA*

BV 2

3.07

2.59

3.07

2.69

2 2* N o t e : T h e a are m e a s u r e d v a l u e s in this l a t t i c e . T h e B are a p p r o x i m a t e

v a l u e s only s i n c e they w e r e m e a s u r e d in u n p e r t u r b e d l a t t i c e s ^ * ' .

TABLE 2

SUMMARY OF NORMALIZED COPPER ACTIVITIES

LOCATION

Fuel

Fuel

Fuel

Fuel

Ring C average

Fuel

Fuel

Fuel

Ring D average

Fuel Average

* Sheath

Sheath

Sheath

Sheath

Coolant Average.

Calandria Tube Av.

Cell Edge, Thimble

Moderator Av.

Q

INDEX

AAABAC1

AC2

AC

AD1

A03

AD2ADAf

AS,A

AS,B

AS,C1

AS ,D1

ncool ant

ACT

Aedge

Am

NORMAL

AIR COOLANT

0.824

0.844

0.925

0.930

0.928

1 .108

1 .109

1.114

1.110

1 .000

1 .005

1 .026

1.031

1 .028

1.089

1 .417

2.061

2.024

IZED ACTIVITY

HEAVY WATER COOLANT

0.757

0.794

0.914

0.912

0.913

1.144

1 . 140

1.137

1 .140

i .000

1 .020

1 .028

1 .021

1 .029

1 .108

1 .537

2.143

2.096

* Note, these values are the average activity of the strip around a fuelsheath, relative to the average activity of the fuel of that element.

T A B L E 3

D E T A I L E D N O R M A L I Z E D C O P P E R A C T I V I T I E S

Loca t ion

CT*

M'V

1-1

M

M

M

M

M

M

M .

H

M

M

M

M .

M :

M :

M :

: N

: S

: E

: W

: E

: E

E

E

E

E

E

E

E

E

E

E

E

N-S

N-S

N-S

N-S

Radius(cm)

6.69

6.69

6.69

6.69

6.82

7.37

8.07

9.02

9.99

10.97

11.97

12.97

13.97

15.00

15.99

16.99

17.99

14.53

14.92

15.56

16.42

Normal ized A c t i v i t y

Ai r Cool an t

1.415

1.419

1 .425

1 .408

1 .421

1.557

1 .678

1 .807

1 .895

1.961

2.020

2.052

2.062

2.055

2.030

1 .972

1 .940

2.076

2,092

2.138

2.216

1 Heavy Water

1 .529

1 .544

1 .545

1 .532

1 .548

1 .654

1 .778

1 .892

1 .9842.049

2.106

2.134

2.143

2.152

2.111

2.064

2.004

2.147

2.173

2.220

2.271

Loca t i on

M

M

M

M

M

M

M

M

M

M

M

M

M

M

M

M

M

M

M :

M :

M :

: N-S

. N-S

: N-S

: N-E

: N-E

• N-E

N-E

N-E

N-E

N-E

N-E

N-E

N-E

N-E

N-E

N-E

N-E

N-E

N-E

N-E

N-E

Radius(cm)

17.50

18.71

20.08

6.82

7.28

7.98

3.99

9.97

10.97

11.97

12.97

13.97

14.97

15.97

16.97

17.97

18.95

19.36

20.95

21 .95

22.95

Normali zed A c t i v i t y

A i r Coolant

2.262

2.295

2.314

1 .436

1 .541

1 .669

1 .827

1 .932

2.031

2.097

2.161

2.217

2.256

2.276

2.285

2.3042.321

2.313

2.322

2.289

2.294

Heavv Water Coolant

2.319

2.355

2.352

1 .556

1 .643

1 .783

1 .906

2.012

2.109

2.172

2.233

2.292

2.310

2.351

2.369

2.388

2.365

2.376

2.368

2.362

2.310

* CT - Calandria tube•*" M - Moderator

TABLE 4

FOIL RELATIVE ACTIVITIES AND RATIOS

CELL LOCATION

Element A

Element B

Element Cl

Element C2

Ring C Average

Element Dl

Element D2

Element D3

Ring D Average

Fuel Average

Cdlandria Tube N

Calandria Tube S

Cell Edge,Thimble T

Reference

In

1 .862

1 .897

2.009

2.005

2.007

2.233

2.234

2. 254

2.240

2.099

2.644

2.642

3.480

1 .000

Mn

1 .010

1.036

1 .145

1 .154

1 .149

1 .363

1 .358

1 .371

1 .364

1 .232

1.746

1 .748

2.548

1 .000

AIR C00I

Lu

1 .302

1 .317

1 .41;

1 .437

1 .427

1 .631

1 .633

1 .636

1 .633

1 .506

1 .957

1 .954

2.659

1 .000

ANTRInRMn

1 .843

1 .83c1

1 .754

1 .738

1 .746

1 .638

1 .645

1 .644

1 .642

1 .704

1.514

1.511

1 .366

1 .000

RMn

1 .289

1 .272

1 .237

1 .246

1 .242

1.197

1 .202

1.193

1 .197

1.223

1 .121

1.118

1 .044

1 .000

In

1 ,680

1 . 748

1 .901

1 .904

1 .903

2.218

2.224

2.244

2.229

2.030

2.746

2.771

3.509

1 .000

0.

0.

1 .

1 ,

1 .

1 .

1 .

1 .1 .

1 .

1 .

1 .

2.

1 .

HEAVY

Mn

931

983

1 30

128

129

413

413

423416

240

894

889

581

000

WATER

Lu

1 .200

1 .251

1 .391

1 .398

1 . 395

1 .650

1 .657

1 .669

1 .659

1 .494

2.099

2.080

2.697

1 .000

COOLANT

1

1

1

1

1

1

11

1

1

1

1

1

1

pinRMn

.805

. 779

.683

.688

.685

.570

.574

. 577

.574

.638

.450

.467

.360

.000

1

1

1

1

1

1

1

11

1

1

1

1

1

RLuKMn

.290

.273

.232

.240

.235

. 168

.173

. 173

.171

.206

.108

.101

.045

.000

TABLE 5

WESTCOTT SPECTRUM PARAMETERS

CELL LOCATION

Element A

Element B

Element Cl

Element Cl

Ring C Average

Element Dl

Element D2

Element D3

Ring D Average

Fuel Average

C a l a n d r i a Tube N

C a l a n d r i a Tube S

C e l l Edge,Th imb le T

THERMAL PIT DATA

[ T J p (" TD2O>

Lr/VVP

r

0 .0455

0.0452

0 .0415

0.0405

0 .0410

0.03570.0360

0 .0361

0.0359

0.0389

0 .0298

0 .0297

0 .0221

2 1 . 7 5 ° C

3 . 8 7 x 1 0 " 4

AIR COOLANT

rvT~7T~n o

0.0517

0.051C

0 .0461

0 .0451

0.0456

0 .0390

0.0394

0 .0393

0 .0392

0 .0430

0 .0314

0 .0313

0 .0225

AT*

85

79

68

70

69

55

56

54

55

63

3?

32

11

HEAVY

r

0.0432

0.0421

0 .0375

0.0376

0.0375

0.0321

0.0323

0.0324

0 .0323

0.0354

0.0260

0.0271

0.0215

21,83°C

2 .93 x 1 0 " 4

WATER COOLANT

r/T7T"n o

0 .0490

0.0474

0 .0415

0.0418

0.0417

0 .0346

0.0349

0.0350

0 .0348

0.0387

0 .0273

0.0284

0 .0220

AT *

84

79

65

67

66

46

47

47

47

57

29

27

11

= [T

T A B L E 6

V A L U E S OF i , C , AND R E L A T I V E U 2 3 5 F I S S I O N RATES

CELL LOCATION

Element A

Element B

:1ement Cl

Element C2

ting C Average

Element DlElement D2

Element D3

Ring D Average

:uel Average

THERMAL PIT DATA

TD20

RCci

nf i ssionsRCd

HEAVY

6

0.0768

0.0719

0.0603

0.0581

0.0592

0.0421

0 .04080.0405

0.0411

0.0511

21.83°C

141

3100

WATER COOLANT

C

1 .4936

1 .4681

1 .4296

1 .4212

1 .4254

1 .40001 .39121 .3864

1.3925

1.4138

U 2 3 5 Fissions

0.746

0. 787

0.902

0.913

0.907

1.133

1.1461.162

1.147

1 .000

I

0.0810

0.0770

0.0672

0.0655

0.0664

0.0484

0.04820.0477

0.0481

0.0583

21.75°C

108

- 4700

AIR

1

1

1

1

1

11

1

1

1

COOLANT

C U 2 3 5

. 3909

.3590

.3719

.3535

.3627

.3846

. 3900

.3893

.3880

.3765

Fission

0.814

0.842

0.921

0.933

0.927

1 . 105

1.1101 .120

1.112

1 .000

L WIMS SPECIFICATIONS

-; K1.1 K IS AT * ! . . • a v . - l t . D?0 OOI."HT, ° I J * ' b / ; <

SE ) IF'-i-F. >.

NO?1 JP.1.4 JNR.) IS 37 - 6 35 7 4 J 1NMFSi 33N S F i t T t J J 4

•IMATPRIAL 1? 0•JPF\CT ft

CA5F DATA

I N I T I A T E

1 1 . 1 0 * ' . ' M t > 2 q 4 . a ' ) 0 0 ( l 1 n 0 3 2 0 0 2 / J . ( l ' i 1 / V i i .0 6 0 0 1 0 , n p q J() 7 77>Hg104'- f>lQ U4.H00Oo0O0 4 2»02 ,;»•!) 1 17320 .0 6001 n.' l?'5U077701Q

lu4 i-7>,if) I 0 11)" A T F i i m . 3 t .->

AN'J'I|_U

T A B L E 8

W I M S / E X P E R I M E N T A L R A T I O S

PARAMETER

A

c

u 2 3 5Fission

Rate

COOLANT

AIR

HeavyWater

AIR

HeavyWater

AIR

HeavyWater

RATIO

DSN/Expt.

PIJ/Expt.

DSN/Expt.

PIJ/Expt.

DSN/Expt.

PIJ/Expt.

DSN/Expl.

PIJ/Expt.

DSN/Expt.

PIJ/Expt.

DSN/Expt.

PIJ/Expt.

R I N G A V E R A G E V A L U E SA B C D

1.089 1.060 1.017 0.970

0.967 1.000 0.994 0.966

1.077 1.045 1.017 0.968

0.962 0.983 1.000 0.978

1.000 1.010 0.980 0.994

0.991 1.010 0.982 0.996

1.002 1.001 0.987 0.995

0.994 0.997 0.987 0.995

0.953 0.979 0.995 1.011

1.003 0.989 0.996 1.005

0.974 0.985 0.996 1.006

1.013 1.004 0.997 1.000

FUELAVERAGE

1 .002

0.980

0.999

0.986

0.992

0.993

0.993

0.993

1 .000

1 .000

1 .000

1 .COO

TABLE 9

WIMS/EXPERIMENTAL RATIOS

PARAMETER

C u 6 3

React ionRate

RLu

Mn

COOLANT

Ai r

HeavyWater

Ai r

HeavyWater

RATIO

DSN/Expt.

P U / E x p t .

DSN/Expt.

P I J / E x p t .

DSN/Expt.

P U / E x p t .

DSN/Expt.

P U / E x p t .

RING AVERAGE VALUESA B C D

0 . 9 5 2 0 . 9 7 8 0 . 9 9 8 1 . 0 0 8

0 . 9 9 9 0 . 9 9 4 0 . 9 9 8 1 . 0 0 2

0 . 9 7 3 0 . 9 8 6 0 . 9 9 4 1 . 0 0 7

1 . 0 1 0 1 . 0 0 5 0 . 9 9 5 1 . 0 0 1

1 . 0 1 1 1 . 0 1 3 1 . 0 1 2 1 . 0 0 6

0 . 9 9 5 1 . 0 0 5 1 . 0 0 7 1 . 0 0 3

0 . 9 9 7 0 . 9 9 7 1 . 0 0 0 1 . 0 1 1

0 . 9 8 6 0 . 9 9 1 0 . 9 9 8 1 . 0 0 9

FUELA V .

1 . 0 0 0

1 . 0 0 0

1 . 0 0 0

1 . 0 0 0

1 . 0 0 9

1 . 0 0 4

1 . 0 0 5

1 . 0 0 2

COOLANTA V .

1 . 0 3 3

1 . 0 2 2

1 . 0 3 8

1 . 0 2 5

C . T . -

1 . 0 3 3

1 . 0 1 6

1 . 0 3 4

1 . 0 1 0

1 . 0 0 4

1 . 0 0 0

0 . 9 9 6

0 . 9 9 5

MODERATORA V .

1 . 0 4 0

1 . 0 0 2

1 . 0 4 0

1 . 0 0 2

CELL +

EDGE

1 . 0 1 1

1 . 0 1 3

1 . 0 0 7

1 . 0 0 9

* Calandri a Tube

+ WIMS value is foe radius of 14.29 cm.

O 37" ELEMENT BRUCE U 0 £ FUEL

• 19-ELEMENT U METAL FUEL

ASSEMBLY CONTAININGDEMOUNTABLE BUNDLE

x FOIL THIMBLE

• O O O O O 4

o o o o o o o *o o o o o o o o oo o o o o o o o oo o o O O O0 0 0 0 0 * 0 0 0

THERMALPIT REGION

O O \ O O O 0 O O O

NORTH

ZED" 2CALANDRIA WALL

FIG. I: TYPICAL LATTICEIN ZED-2

ARRANGEMENT

AL SHEATH U-METAL FUEL

I.D.O.D.

1.38 cm1.59cm

DIAMETERDENSITY

1.31 cmI8.93g/cm3

COOLANT SPACE

AL COOLANT TUBE

I.D.O.D.

8.58 cm8. 89 cm

FIG. 2 : CROSS SECTION THROUGH 19-ELEMENTU METAL FUEL ASSEMBLY.

Zr-4 SHEATHI.D. 1.22cmO.D. 1.31 cm

Al PRESSURETUBE ^

I.D. 10.39 cmO.D. 11.02 cm

U02 FUELDIAMETER T.2I cmDENSITY IO.5Og/cm

AIR ANNULUSAl CALANDRIA TUBEID 12.70 cmO.D. 13.34 cm

FIG. 3: CROSS SECTION THROUGH37-ELEMENT FUEL ASSEMBLY

< < < z «

UJ_ loz300

_lUJ

u_

UJo3trm

i-zUJ

Ul_lUJ

I

rO

r̂ID

- 5:DEMOUNTABLE BUNDLE

COMPONENTS

NATURAL URANIUM METALFOIL (BARE)*

DEPLETED(WRAPPED)

ALUMINUM CAN

U02 PELLET

In^AI, Lu-Mn-AI FOILS (WRAPPED)

'CATCHER FOILS NOT ILLUSTRATED

FIG. 6 : DEMOUNTABLE PACKAGE AND COMPONENTS

FOILS BETWEENFUEL PELLETS

CELL CORNER

ALUMINUM

FRAMEWORK

-COPPER WIRE

CELL EDGE

FOIL

FIG- 7: SHOWING DETECTOR LOCATIONS

NORTH

FIG. 8 SECTOR FOIL CUTTING SCHEME

NORMALIZED COPPER ACTIVITY

NORMALIZED COPPER ACTIVITY

JOAD

IUS

n

o-

IM -

*--

a> •

co

ro

Oro

Oro

co

ro

0.80

1.00

1.20

• i i • i i t

>

1°1"

5'8

1

> i* $

1.60

1.40

i i i

AIRC

00ILA

NT

1

WE

RA

GE

(:O

PPEV

ITIES;

P

•n

cm

•n

0193

2

z

NORMALIZED COPPER ACTIVITY

NORMALIZED COPPER ACTIVITY

oo

o o— i —

I

CDm

-o31mCOCO

>JO

as

c:

3 -

ro

OoJO

ro

M

FIG. 13: COPPER ACTIVITYAROUND A FUEL PIN ."HEAVY WATER COOLANT

JTO- —~* r L I

X/

\J>

H

\

\

(TIT•*" LLL

V( /^

wahMA

Ii

V -r \

11

•\ r

1i

•i .̂ ^1 ELEMENT

\

ZED COPi .

7^/y ^~<

X\

\aer

2T0'

90*

270"

0"

^ ^ X X \ M A \ U S

\

mfw

friWfflinliLLJi

1 / / /

7 2SO?

, /X

X' yK

-.) ELEMENT 01 \

T^-^lff-^J/NOHMAUIED COPPEflf-l^.

\vc\ 4 .X180*

FIG. 14. COPPER ACTIVITYA FUEL PIN: AIRCOOLANT

270"

IBO"

FIG. 15: ARRANGEMENT OF URANIUM FOILS

PAIRS OFTHIN ALUMINUMCATCHER FOILS

THIN ALUMINUMCATCHER FOILS

28.5 mm

- BARE NATURALURANIUM FOIL

-DEPLETED URANIUMFOIL WRAPPED INTHIN ALUMINUM

FIG. 16: P(t) VERSUS TIME*

P(t)

1.26

1.22

1.18

1.14

"AFTER A. OKAZAKI

1.10-

55 40 55 §5TIME AFTER SHUTDOWN (UNITS IO3s)

10

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ISSN 0067-0367

has been assigned to this series of reports.

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