ÅE-299 UDC 621.039.512.2

621.039.543.4

The Measurement of Epithermal-to-Therma! o

U-238 Neutron Capture Rate (p28) in Ågesta

Power Reactor Fuel

G. Bernander

AKTIEBOLAGET ATOMENERGI

STOCKHOLM, SWEDEN 1967

AB-299

THE MEASUREMENT OF EPITHERMAL-TO- THERMAL

U-238 NEUTRON CAPTURE RATE ( o _ ) IN ÅGESTA Zo — — — — — —

POWER REACTOR FUEL

G. Bernander (ASEA)

SUMMARY

The ep i the rma l - to - the rma l neutron capture ra te ra t io p_„ in

U-238 in Ågesta fuel has been measu red by the chemical separat ion

method. The method involves the isolation of Np-239 from uranium

and fission products by r eve r sed phase part i t ion chromatography.

Although somewhat e labora te , and in spite of difficulties with res idua l

fission products , the method has yielded reasonably accura te r e s u l t s .

Fu r the r development work on chemical p rocedures may lead to some

improvement . A comparison with the coincidence method - e lectronic

separat ion of act ivi t ies - has not shown any la rge sys temat ic differen­

ces between the two methods.

The separat ion of the epi thermal U-235 activation from the

total has been achieved by means of the " l / v subtract ion technique"

using copper foils as the l / v moni tor . The complementary the rmal

column i r radia t ions required have been performed in the r e s e a r c h

r eac to r s TRIGA (Helsinki) and Rl (Stockholm).

F r o m the measured p_„ values the resonance escape probabi­

lity (p) and the initial conversion ra t io (ICR) may be calculated using

c ros s - sec t ion data and other lat t ice p a r a m e t e r s . Comparisons with

theoret ica l values of p and ICR as calculated with the BURNUP

latt ice pa ramete r code a re favourable. The re su l t s for the 19-pin

c lus ter of the Ågesta fuel a re summarized below.

Tem­

pera­

tu re

<°0

35

212

E x p e r i m e n t

P28

0.365 + 0.009

0.422 + 0.015

p

0.902 + 0.003

0.882 + 0.004

ICR

0.775 + 0.005

0.833 + 0.008

T h e o r y

P

0.892

0.879

ICR

0.795 ( 20°C)

0.838 (220°C)

Prin ted and distr ibuted in September 1967

LIST OF CONTENTS

Page

1. Introduction 3

2. Derivation of p_„ from measured quantities 4

3. Exper iments 10

3. 1. Descript ion of fuel and i r radia t ion technique 10

3 .2 . Measurements 12

3 . 3 . Np-239 activity determinat ion 14

3 .4 . Auxiliary measurement s 19

3 . 5 . Cu foil activity counting 22

4. Resul ts 23

4 . 1 . Activation data 23

4 . 2 . Derivation of p ? f i 25

4 . 3 . Calculation of p and ICR 29

5. Summary and discussion 32

Acknowledgements 33

References 34

List of figures 36

- 3 -

1. INTRODUCTION

In conjunction with the commissioning physics t e s t s in the Åge sta

power reac to r [ 1 , 2 , 3 ] the r e sonance - to - t he rma l U-238 neutron capture

r a t e r a t io , p?ft» w a s measured together with severa l other p a r a m e t e r s .

This was par t of an effort to a s s e s s the resonance absorption cha rac ­

t e r i s t i c s of heavy water modera ted and cooled uran ium oxide la t t ices at

both ambient and operat ional t e m p e r a t u r e s . These c h a r a c t e r i s t i c s , as

r epresen ted by p?o» the initial conversion ra t io (ICR), or the resonance

escape probability (p), a re important in predicting initial excess r e a c ­

tivity and fuel burnup.

A chemical method was adopted to separa te the fission products

f rom the induced Np-239 act ivi t ies that were taken as a m e a s u r e of the

neutron capture ra te in U-238. Thomasen and Windsor [4] repor ted

favourable resu l t s with a chemical separat ion technique, and the high

fluxes available in the Ågesta r eac to r were an advantage with this

method since the high specific act ivi t ies possible would eliminate the

uranium background problem.

The " l / v subtraction method", originally proposed by Egiazarov

et a l . [ 5 ] , was used to separa te the epi thermal ly induced Np-239

activity f rom the tota l . In this method the l / v activation is monitored

with a separa te l / v de tec tor , for which copper foils were employed.

In order to obtain absolute values of capture ra te r a t i o s , an auxil iary

i r radia t ion must be ca r r i ed out in an essent ia l ly pure the rmal flux.

The fuel samples were 10 m m long UO~ pellets with the same

diameter as the ordinary fuel. After i r rad ia t ion , the pellets were

dissolved and a small fraction of the solution was processed to sepa­

ra te the neptunium from the uranium and fission products . The

neptunium finally collected and to be counted with an Nal(Tl) c rys t a l

was thus in a dissolved s ta te .

There a r e some advantages in using large fuel samples and

subsequent dissolution instead of counting very thin foils of fuel ma te ­

r i a l . No i r radia t ion geometry difficulties a r i s e since modera te su r ­

face defects and slight misal ignments or gaps between pellets introduce

only very small e r r o r s . The dissolution yields an averaged and uniform

activity distr ibution in the counting samples , reducing also counting

- 4 -

geometry p rob lems . The relat ively high i r radia t ion flux neces sa ry in

the Ågesta i r rad ia t ions (because of other simultaneous measuremen t s )

a lso favoured the dissolution technique, which allows an a r b i t r a r y activity

fraction in a sample to be counted. The high specific activity at the same

t ime reduces problems of na tura l uranium activity background.

A disadvantage of chemical separation work is that exceptionally

grea t care must be exerc ised in order to avoid excessive activity losses

in the various processing s tages . The procedures were a lso ra ther

e laborate and t ime consuming, especially since the separat ion yield had

not been quite complete, necessi ta t ing further complementary m e a s u r e ­

ments to establ ish correc t ion fac to r s .

In one of the three measurements repor ted h e r e , a comparison

was made with coincidence counting on unseparated dissolved samples .

In this case no significant sys temat ic differences between the two

methods could be inferred.

In the following, the theore t ica l background is f i rs t t rea ted

(section 2); then the experiment and the procedures a r e descr ibed

(section 3). Section 4 gives the experimental r e s u l t s , and a summary

and discussion concludes the paper .

2. DERIVATION OF p 2 g FROM MEASURED QUANTITIES

In order to emphasize the importance of p ? o, the express ions for

resonance escape probability (p) and initial conversion ra t io (ICR) as a

function of p?„ will f i rs t be developed; then p?o will be defined in t e r m s

of measured quantit ies.

The definitions of p and p ? 8 found in the l i t e ra ture vary somewhat

depending on the lat t ice pa ramete r model and the form of c ros s - sec t ion

convention used. In this paper we shall conform to the Swedish lat t ice

pa rame te r recipe as represented by the BURNUP code [6] which

employs Westcott-type c ross - sec t ion data. Thus p , as calculated from

resonance integral data, involves escape from U-238 resonance absorp­

tion only. F a s t neutron capture in U-238 (above fission threshold) is

wholly accounted for in the fast fission factor , whereas all l / v absorption,

including that in the resonance region, is contained in the the rmal

utilization factor.

- 5 -

The following notations will be used for the quantities occurring

in the formulae:

N = neutron radiative capture in U-238

M = neutron absorption (including fission) in U-235

Indices t, r and s denote "thermal" (all l/v capture, re­

sonance, and fast absorption respectively

P , P and PA are the fast, intermediate, and thermal non-s r t

leakage probabilities

f = thermal utilization factor

T| = number of fast neutrons produced from U-235 fission

per thermal absorption in fuel

Q = fast fission factor

v = number of fast neutrons formed per fission

£ = macroscopic absorption cross-section

Indices a, c and f denote total absorption, radiative

capture, and fission respectively

L, = fast neutron group migration length

B2 = the buckling of a critical system

F = fission rate

a = capture-to-fission cross-section ratio

R = total-to-thermal neutron capture ratio in U-238

C = neutron capture in Cu-63

Resonance escape probability (p)

The p-factor is defined as the fraction of the total number of

fast neutrons available for slowing down past the U-238 fission threshold

that escape capture in the U-238 resonances (excluding the l/v con­

tribution) during moderation. Then the fraction absorbed in the U-238

resonances is approximately given by (all U-235 capture is accounted

for as thermal)

N /P 1 " P = (Nt+M)/PsPrPtpf W

Since the effective multiplication constant may be written as

6 -

and

k , , = k • P P P = T] e pf • P P P. eff t» s r t i - r s r t

N t + M = T-(28T ' N t 3. t

(whe re £ (U) = £ , (28) + £ (25)) eq . (1) t a k e s the f o r m cl ctX 3.

keff S a t < 2 8 > N r (?)

In an inf ini te s y s t e m k , r / P = k , w h e r e a s in a f ini te ju s t ' e t i ' s <»

c r i t i c a l s y s t e m k , , = 1 and P = l / ( l + L B ) .

Now, p ? o i s def ined by

N - N t N + N t _ r s , , , p 28 _ N N V>

so tha t

N N

N ^ = p 28 " N^* ( 4)

w h e r e the t e r m N / N a c c o u n t s for the f a s t c a p t u r e in U-238 inc luded

in p?Q« If the U-235 and U-238 t o t a l f i s s i o n r a t e s a r e F ? ( - and F_„

r e s p e c t i v e l y , t hen

S _ ( 2 8 ) N = — ^ 7 r • F = (v F

s S . (28) * 2 8 ff28 28 f s

and

N = a t

t ' S f(25) ' * 25

Combining t h e s e two quan t i t i e s one o b t a i n s :

- 7 -

N g £ f ( 2 5 ) F ^ g

K = "28 * a 7 ^ ' F25 U

T h u s , if F n o / F 1 c i s m e a s u r e d or i s c a l c u l a t e d f r o m the f a s t f i s s i o n ' 28 ' 25

f a c t o r , if known, a c c o r d i n g to

F 2 8 , .. V 2 8 " 1 " Q ? 2 8 ( e -1 )

F 2 5 V25

t h e n the f a s t c a p t u r e c o n t r i b u t i o n m a y be e s t i m a t e d .

Combin ing e q . (2), (4) and (5) and us ing the r e l a t i o n

S f (25) Tl

^ W V25

one g e t s i n s t e a d of eq . (2):

Ssff S a t ( 2 8 ) k e f f a 2 8 F 2 8 k

10 T j a t v w / " " " v 2 5 " s " 2 5 p =

TIP: * "Our * p28 - I^TP- • F- <6>

The l a s t t e r m in (6) c o r r e c t s for the f a s t c a p t u r e i nc luded in the

m e a s u r e m e n t of p?o« T h i s r e l a t i o n i s equ iva l en t to the one d e r i v e d

for p by Kouts and Sher [ 7 ] .

In i t i a l c o n v e r s i o n r a t i o (ICR)

The ICR i s def ined a s the n u m b e r of f i s s i o n a b l e n u c l e i f o r m e d

f r o m n e u t r o n c a p t u r e in f e r t i l e m a t e r i a l p e r f i s s i o n a b l e n u c l e u s con ­

s u m e d in f r e s h fuel . H e n c e , for u r a n i u m fuel (neg lec t ing c a p t u r e in

U-234)

T/--D _ Nonf i s s ion c a p t u r e r a t e in U-238 _ N T o t a l c a p t u r e r a t e in U-235 M

or

N N t N t I C R = N;-u = ( 1 + p 2 8 ) - i M

or

- 8 -

ICR = ( l + p 2 8 ) . f ^ j (7) av

This relation will be used for calculating ICR from (l + p~R) which is

measured. For comparison with theory the "modified relative con­

version ratio" (RCR*) is frequently used:

* _ <N/F25>lat RCR* = (N/F-J 25'th

where the U-235 fission rate F?c. is determined from a measurement of

the fission product activity. The index "th" implies a pure thermal

flux. Then the ICR is obtained from

The RCR fi may be expressed also in terms of P-?Q« From eq. (7) and

(8) it is easily shown that

Derivation of p^q from measured quantities

p?o has already been defined in eq. (3) as the ratio between re­

sonance (excluding l/v) plus fast capture and the total l /v capture in

U-238. This definition is a consequence of using Westcott cross-

section formalism in lattice parameter calculations. Therefore, the

cadmium ratio concept is not introduced. One has then:

N + N N - N r s t _N_ .

p28 = N " N. = N. t t t

The quantity directly measured is

R = N 1 = 1 + p 2 8 <1 0)

- 9 -

F o r the purpose of subtracting the l /v capture (N.) f rom the

total capture obtained when measur ing the Np-239 induced act ivi ty, an

auxi l iary l / v detector is requi red (Cu foils). If the l / v capture in the

Cu foils is denoted by C , eq. (10) may be writ ten as

N S N ^ t ,,,, R = S ' * T S *? (u)

where N/C is the capture ra te ra t io in the latt ice and C /N. = C*/N*

may be determined from a t he rma l column i r radia t ion (as te r i sks de­

note t he rma l column values) .

Eq. (11) does not take into account the fact that copper has a

finite resonance in tegra l . Moreover , the the rmal column does have

an epi thermal flux component, although sma l l . Eq. ( l l ) is then modi­

fied as follows. Since

C = C. + C = C. (1+C /C . ) t r t x r ' t '

C* = C* + C* = C*(l+C*/C*) t r t v r ' t '

N* = N* + (N*+N*) = N*(l+N*/Nf+N*/Nf)

eq. (11) is converted to

N/N* d + C r / C t ) ( l + N y N * + N * / N * ) K _ C /C* * (1+C*/C*) ( 1 Z J

where N, N*, C and C* a re determined experimental ly f rom the

Np-239 and Cu-64 ac t iv i t ies . The cor rec t ion factors a r e es t imated

from c ross - sec t ion data and auxil iary measu remen t s on ep i the rmal -

to - the rma l flux r a t i o s . The cor rec t ions a r e t rea ted in more detai l in

section 4. 2.

Even though R may be measured accura te ly , it is seen f rom

eq. (10) that the uncertainty in p_„ may become ra ther large when p ? R

is smal l . Fo r example, if R has an uncertainty of + 1 %, then p_„

will have an e r r o r of near _+ 4 % when p_„ is of the order of 0. 35 (as

- 10 -

in the case of the cold Ågesta la t t ice) . Thus ( l -p) according to eq. (6)

will have an e r r o r in excess of J; 4 % and, when account is taken of un­

cer ta in t ies in the other data, perhaps some + 5 %. However, if p is

approximately 0. 9 (*=» 0. 89 in Ågesta) , then the e r r o r in p will be l ess

than _+ 0.6 %. This would be sat isfactory in view of the ra ther l a rge r

uncer ta int ies in theore t ica l p va lues .

In the measurement of p ? R ca re must be taken to i r rad ia te the

fuel samples in a latt ice region where the neutron spec t rum is unper­

turbed by ref lectors or control r o d s . With Ågesta fuel the spacers

between fuel bundles give r i s e to a t he rma l neutron flux peak and a

corresponding depletion in the epi thermal flux component. This effect

will tend to dec rease p?o» a s a consequence, too low ICR-values and

too high p-values will be obtained in the vicinity of the spacer gaps . In

the experiments the i r rad ia ted fuel samples have not been nea re r to

such a gap than 19 cm, which was thought to be adequate in order to

avoid these effects.

3. EXPERIMENTS

The procedures may be outlined briefly as follows. UO_ pellets

and foils were i r radia ted in the Ågesta reac tor and in a the rmal column

almost simultaneously. After i r radiat ion the samples were t ranspor ted

to the Studsvik labora tor ies where Cu foil counting was s tar ted imme­

diately. The UOp pellets were processed in the radiochemical labora­

tory to obtain separated neptunium samples (solutions) for the counting

of Np-239 activity by means of g a m m a - r a y spectroscopy. The fission

product solutions and separation column fillings were also collected in

order to check Np l o s s e s .

3 . 1 . Description of fuel and i r radiat ion technique

The fuel in the Ågesta lattice is arranged in a 27. 0 cm square

latt ice (fig. 1) corresponding to a modera to r - to -UO ? volume ra t io of

15.6 (including coolant). The experimental fuel assembly was i r radia ted

in a cent ra l position (C 10 in the notation of fig. 1) where the neutron

flux would be essent ial ly unperturbed by the control rods needed to shim

- l i ­

the excess react iv i ty . The axial flux profile is exemplified in fig. 2 as

measu red with a copper wire positioned on the outside surface of the

coolant channel. The spacer gap flux peaking effect is c lea r ly shown.

The fuel a s sembl ies each compr i se four 19-pin bundles con­

nected end-to-end and contained in a coolant tube, see fig. 3a. The

total active fuel length is 305 cm including the three spacer gaps , which

a r e 4. 5 to 4. 8 cm each depending on the volume of the expansion space

in each pin. The UO_ length in each bundle, then, is 72.5 to 72.8 cm.

All s t ruc tura l ma te r i a l s a r e of Zi rca loy. The c ros s - sec t i ona l di-

mensions a re given in fig. 4a. The UO_ density was 10. 6 g / c m .

The special exper imental fuel a s sembly was made f rom an ordi ­

nary assembly by modifying it to include a removable group of six r e ­

presentat ive fuel pins in one of the four sections (second from bottom

end). The pins a re normal ly screwed into each other axially, but in

the modified section the exper imenta l pins could be detached la te ra l ly

through a rectangular opening in the coolant tube, see fig. 3b. The

curved plate cut out from the tube was fitted with a top and a bottom

gr id , thus forming a pin holder . When inser ted , the holder was

secured with two lock bol t s . The integri ty of this a r rangement was

sufficient to withstand normal coolant flow conditions at high t empe­

r a tu re (220 C) and p r e s s u r e (34 b a r s ) , and permit ted handling in the

normal manner by the loading machine . The pin holder could be r e ­

moved using special shields and tools .

The special pins containing the exper imental fuel pellets and

copper foils (together with var ious other foils for flux distr ibution and

spec t rum measurements ) were welded tight pr ior to inser t ion into the

exper imental assembly . After i r rad ia t ion , samples and foils were

removed by sawing off the end plugs of the pins .

After i r radia t ion there was a cer ta in delay before the pins could

be removed from the exper imental a ssembly . With the low t empera tu re

i r radia t ion it was necessa ry to wait for the fission product activity to

decay to an acceptable level - a surface dose ra te of some 30 r / h with

the shielding arrangement"used. With an i r radia t ion level of some

10 n / c m s for 90 minutes this meant a seven h o u r s ' waiting t ime .

After a high tempera ture i r radiat ion the limiting factor was the cooling-

down t ime and fuel discharge t ime , altogether some 15 hours .

- 12 -

In the t he rma l column i r radia t ions the fuel specimen has to be of

l imited size since it is a source of fast neu t rons . However, using Cu

foils as l / v detec tor , the condition of equal flux distr ibution in Cu and

adjacent fuel pellet sample is fulfilled only with an extended piece of fuel.

Even quite thin samples in an isotropic neutron flux experience a flux

depress ion that is difficult to determine with adequate accuracy .

3 .2 . Measurements

In al l , th ree measu remen t s were made according to table 1 below.

The second h igh- tempera ture measuremen t was ca r r i ed out at an average

fuel burnup of 0.23 MWd/kg of uranium. However, the exper imental

assembly was sti l l of fresh fuel.

Table 1. List of measuremen t s and i r radia t ion data

Measure­ment

No.

I

II

III3 )

Temperature

o ( C)

35+1

212.5+1

212+1

Sample posi-1) . tion in

experimental fuel assembly (C10) (cm)

20.9

18.9

35.5

2) Control rod insert ion

Control rods

(cm)

A, B , C, D22

C40

A, B, C, D44

A, B , C, D26

A62

A, B, C, D44

A, B , C, D26

A62

A, C40; B , D04

A, B, C, D44

A, B, C, D26

230

91

0

0

<*>50

49

0

92

199

0

0

Distance of sample to nearest con­t ro l rod

(cm)

95

116

95

Thermal column i r radia t ion in

TRIGA (institute of Technology, Helsinki, Finland)

Rl (Stockholm)

Rl (Stockholm)

1) Distance from center of sample pellet to bottom plug in pin (U0 length in each pin is about 72.5 cm)

2) Position designation according to fig. 1. The rod insertion is counted from "fully in" (0 cm);

at 300 cm control rods are wholly withdrawn - such rods are not listed in the table,

3) This experiment was carried out at an average burnup of 0.23 MWd/kgU

- 13 -

The UO? pellets used in the exper iments (sample pel lets and

adjacent pellets) were selected f rom a large batch, taking ca re to avoid

specimens with surface defects . All pel lets in the batch had had their

ends ground perfectly plane and at r ight angles to the axis (length

variat ion within J; 0.03 m m ) , and their d i amete r s were 17.00_+0. 01 m m

obtained through cen te r less grinding. By this means any uncer ta in t ies

depending upon surface defects or gaps between pellets a re quite smal l .

By using 10 and 4 m m long sample pellets in the lat t ice and t he rma l

column respect ively , the effect of misal ignments should also be negl i ­

gible. In the the rma l column the smal le r size pellet was des i red in

order to l imit rod size and because of the relat ively l a rge neutron flux

gradient . Of course , in this case with a near pure t he rma l flux, su r ­

face defects a re not as important as in the la t t ice .

The sample geometr ies a r e shown in fig. 4 and 5. Altogether

five pellets would be i r rad ia ted in each measu remen t ; four in the ex­

per imenta l fuel assembly - positions 1, 2, 4 and 5 in fig. 4a - and one

in the the rmal column. In the lat t ice rods the copper foils were not

positioned next to the sample pellets (see fig. 4b), so that resonance

flux s t reaming effects would be avoided. Since the flux gradient was

quite smal l (compare fig. 2) the Cu activity at the center of the pellet

could readi ly be obtained by interpolat ion. In the the rma l column rod,

however, the Cu foils were placed next to the sample pellet (fig. 5).

Here resonance flux s t reaming is of little impor tance , the t he rma l flux

gradient is l a rge , and it is n e c e s s a r y to avoid end effects on the radia l

flux distr ibut ion.

The copper foils were of the same d iamete r , 17.0 m m , as the

UO_ pellets and 0. 10 m m thick. They were protected f rom catching

fission products by sandwiching them between thin aluminium or s teel

foils 0. 05 m m thick.

The Åge sta i r radia t ion was made in a flux of the order of

10 n / c m s for 90 minutes . The reac tor power was some 10 to 20 kW

but, because coolant circulation flow was the same as at full power

(65 MW), no significant excess fuel t empera tu re above that of coolant

or modera tor was expected. In the high t empera tu re i r rad ia t ions the

coolant and modera tor were heated external ly by en e lec t r ic heater and

could be maintained at a constant t empera tu re (within _+ 1 C). When

- 14 -

poss ible , the power level was recorded automatically by a digital data

acquisition sys tem (RAMSES); otherwise the power was read off and r e ­

corded manually.

The the rma l column i r radia t ions were ca r r i ed out within a few

hours of the Ågesta latt ice i r rad ia t ion . A thermal neutron flux as close

as possible to that in the lat t ice was aimed for; however, this was not

always successful due to the power level uncertainty in the Ågesta r e a c ­

to r . In the Rl reac to r the rmal column, the sample was positioned at

12. 5 cm from the inner column boundary. Since the graphite ref lector

is 90 cm, a total of 102 cm of graphite separated the sample f rom the

reac to r tank. The D-O reflector of the R l core is 5 to 10 cm.

The sample position in the TRIGA reac to r the rmal column gave

about 127 cm of graphite between the sample and the c o r e . Thus the

epi thermal flux component in this case should not be g rea t e r than in the

Rl case . The the rmal flux gradient in the the rma l column sample

positions was 4 .8 %/cm in TRIGA and 3.1 %/ cm in R l .

3 . 3 . Np-239 activity determination

Chemical separat ion

The method of separat ion of neptunium from uranium and fission

products (F . P . ) will be descr ibed briefly. A more detailed review of

the chemical work is found in ref. [8 ] by S . - E . Kroon, who ca r r i ed out

the chemical work and developed the procedures in collaboration with

Å. Hultgren, Studsvik. Basical ly , however, the method is equivalent to

that developed at Kjeller , Norway, by Thomasen and Windsor [ 4 ] ,

The principle of the method is the so-cal led r eve r sed -phase pa r ­

tition chromatography. An extraction agent (TTA) with which Np(lV)

forms a strong complex is adsorbed on hydrophobic ma te r i a l (specially

t rea ted glass powder) in a column to form a s ta t ionary phase . An

aqueous solution with moderate acidity containing the Np, U and F . P .

is allowed to run through the column. Np is extracted by the stat ionary

phase whereas U and F . P . a r e washed away. Then Np is eluted with

high-concentrat ion acid and collected.

- 15 -

The chemical procedures will now be outlined.

a) Dissolving of UO^

Each UO- pellet is dissolved in hot cone. HNC" . With the pellet

size used this is a mat te r of about two hours . Thus a p r i m a r y

stock solution of typically 100 ml is obtained, of which only a

smal l fraction is needed for the subsequent s t ages . F r o m the

stock solution may be taken samples for F . P . counting or for

Np-239 counting by the coincidence method.

b) Conversion to chloride solution

F r o m the ni t ra te solution obtained in a) is taken a 3 % aliquot in

a beaker . After careful evaporation to d rynes s , cone. HC1 is

added. HNO, now escapes and, after further evaporation and

adding HC1 once m o r e , the solution is t r ans fe r r ed to a 25 ml

f lask. In this chloride solution the Np has a valence of +6.

c) Reduction of Np(Vl) to Np(lV)

Hydroxylaminehydrochloride (NH2OH«HCl) is now added and the

reduction is promoted by submerging the flask in a boiling water

bath for 1 hour. Then the solution is diluted with water to 25 ml

thus obtaining a secondary stock solution containing reduced Np

ready for separat ion.

d) Separation of Np from U and F . P .

An aliquot (0. 8 ml containing near ly 0.1 % of the original UO?

pellet activity) is now allowed to run through the separat ion

column (described below). Np is selectively extracted by the

stat ionary phase so that U and F . P . only a r e collected in a con­

ta iner . A "washing solution" (0. 5M HC1 - 0.1M NH2OH- HCl) is

also used to remove all the U and F . P . Then the Np is eluted

with 6M HCl- 1M HF plus H C l - saturated CJHgOH plus

6M H C l - IM HF and collected in a separate Teflon container

for subsequent activity counting.

e) The column glass powder filling is also collected (to determine

any res idual activity) after drying the columns in an oven.

The separat ion column had an ID of 5 m m in the filling section.

The extraction agent, TTA (2-thenoyltrifluoroacetone) dissolved in

xylene, was absorbed on hydrophobic glass powder (75 - 150 mesh) ,

- 16 -

thus forming the stat ionary phase of the column. The separa t ion capa­

city is l imited with respec t to uranium content in the sample to be

separa ted; for these columns the l imit quantity amounts to some 30 mg.

Accordingly, the specific Np-239 activity must be ra ther high to obtain

good counting s t a t i s t i c s .

Natural ly , the chemical work will involve inevitable los ses of

act ivi t ies in the various s t ages . In one tes t th ree aliquot taken from a

single p r i m a r y stock solution were independently p rocessed to form

three "ident ical" secondary stock solutions (reduced Np). F r o m each

of these solutions three aliquots were separa ted and the activity of the

nine samples of separa ted Np obtained was then counted. The resu l t

indicated that the reproducibil i ty of the chemical p roce s se s was within

_+ 1. 0 % as determined from the spread of r e s u l t s . The averages of the

three separat ions corresponding to each of the three reductions agreed

within 0.25 % ( i . e . well within counting s ta t i s t ics ) , showing that the

par t of the chemical work leading to the secondary stock solution is

sufficiently accura te . Apparently the variat ion in separat ion yield of

Np-239 and/or F . P . res idue accounts for the major uncertainty con­

tr ibut ion. The quality of the separat ion process may be checked by

auxil iary measurement s (see section 3 .4) . La te r , the samples

(aliquots and stock solutions) were carefully weighed instead of mere ly

measur ing their volumes . This led to some improvement since it was

found that weighing was ra ther more accura te .

Whereas the separat ion yield of Np is 99.5 % or m o r e , the

elution of Np from the column is not quite complete, som 2 to 3 % r e ­

maining in it. Since this res idual activity is var iab le , the glass powder

filling of the column is collected and its activity measu red . It has been

found, a lso , that some F . P . follow the Np, notably Te and Z r . To

check this contamination the g a m m a - r a y spect ra of al l Np, F . P . and

glass powder samples were recorded in order to a s s e s s correc t ion

factors (see section 3 .4) .

F o r each of the 5 i r rad ia ted UO~ pel le ts , only one reduction was

ca r r i ed out. However, corresponding to each pellet 2 or 3 separat ions

were made, giving a total of 1 2 - 15 separa t ions .

- 17 -

Np-239 counting

The beta decay of Np-239 into Pu-239 mainly involves the e m i s ­

sion of g a m m a - r a y s with energies of 106 .4 , 228 and 278 keV. The

la t te r two of these a re fa i r ly s trongly converted, giving r i s e to the

103. 7 keV ( K a j and 99- 5 keV (KcyJ X - r a y s in plutonium. These

X - r a y s combine with the 106.4 keV g a m m a - r a y into the 105 keV photo

peak observed in the scintil lation spec t romet ry pulse height spec t rum

of Np-239. F ig . 6 shows the spec t rum obtained with a la rge Nal (Tl)

c rys t a l for both separa ted and unseparated samples and normal ized at

105 keV. In order to suppress efficiently any res idual F . P . that may-

remain after separat ion, the Np-239 activity is counted in an in terval

around the 105 keV peak. The F . P . spec t rum is demonst ra ted in

fig. 7, in which the most prominent peaks a re identified. No uranium

activi ty is noticeable because of the high specific F . P . activity in the

sample .

The Np-239 activity in the liquid samples was measu red with

a 1. 75 in. d iam. and 2 in. thick Nal (Tl) c rys ta l mounted on an

EMI 6097B photomultiplier tube. The counting equipment included a

Nuclear En te rp r i se s non-overloading (NE 5202) l inear pulse ampl i f ier ,

a Landis and Gyr single-channel analyzer (1.2 (j,s resolut ion t ime) ,

and an ACEC decade sca ler (DM160) provided with an electronic t i m e r .

The stabilized high voltage source was also included in the sca le r unit.

F ig . 8 shows the geometr ica l counting a r r angemen t . The

Teflon containers for the Np-solution had been accurate ly machined;

inter comparison of count r a t e s using a Np-23 9 solution did not r evea l

differences in excess of the counting s ta t i s t ics e r r o r , 0.2 %. The

container covers were tightly fastened with adhesive tape to prevent

the liquid from evaporating. In spite of th i s , a slight evaporation was

observed (by weighing), probably due to escape of the ethyl-alcohol .

F ig . 9 depicts the spec t rum of an unseparated sample that was

used to set the high voltage and amplifier gain of the counting equip­

ment . The 105 keV photo peak was positioned between 10 and 15 volts

(maximum pulse height was 100 V). Great ca re was taken to adjust

the analyser channel position (lower d iscr iminator level) and width.

On the one hand, a smal l width will re ject a maximum of res idual

- 18 -

F . P . - in an unseparated sample the F . P . contribution in one instance

var ied f rom 12. 7 % to 7. 3 % according as the channel width was de­

c reased f rom 50 to 30 keV (fixed channel position at 86 keV). On the

other hand, the effect of electronic drift will be more marked using a

na r row "window"; count ra te also will be re la t ively smal l . The p roce­

dure was to select a channel width somewhat l a rge r than the half-

maximum 105 keV line width and then measure the count ra te as a

function of channel position. The count ra te then neces sa r i l y exper i ­

ences a maximum (provided the channel embraces one peak only),

yielding a smal l "plateau" some 0.4 to 0.6 volt wide inside which the

count ra te var ied l ess than 0. 25 %. The effect of var iable channel

width on count ra te was typically less than 0. 9 % per 0.1 volt; however,

the channel width drift apparently lay well within J; 0. 05 V.

Electronic stabili ty was checked in the f i rs t two measu remen t s

with a Th-228 s tandard, the count ra te of which was ve ry sensit ive to

gain var ia t ions , g rea te r than 3 % per 0.1 V. In the las t measurement

an Am-243 standard (half-life 7950 y) was used. Except for the 75 keV

g a m m a - r a y from the alpha decay of Am-243 into Np-239, the spec t rum

is pure Np-239 for a fresh sample . The count ra te sensi t ivi ty was

about 0. 8 % per 0. 1 V (due to 75 keV peak) whereas channel width

var ia t ions should be about the same as for the Np-239 samples , In all

cases the checks indicated sat isfactory stabili ty.

The counting procedure aimed at eliminating any short or long

t e r m electronic dr i f t s , in the meanwhile accumulating a sufficient

number of counts for good s ta t i s t i c s . The samples were counted in ter­

mittently over a span of about two days beginning at about 72 hours

after i r radia t ion . The repeated counting of each sample also reduced

any random e r r o r ar is ing from sample positioning in the counter .

The count r a t e s obtained were correc ted for dead- t ime , back­

ground, and decay. The dead- t ime correct ion was significant only in

the second measurement with a count ra te as high as 2700 c / s e c ; how­

ever , all samples had quite s imi lar ac t iv i t ies , so the dead- t ime un­

certainty became negligible when activity rat ios were formed. The

detector background was always less than 7 c / s ec and the background

induced by the sample - as measured on a separated sample f rom un-

- 19 -

i r rad ia ted uran ium - was only about 1 c / s e c . The dis integrat ion con-- 4 . - 1

stant for Np-239 used in the decay cor rec t ion was 2. 052 • 10 min

(half-life 2.346 days) .

Final ly , the Np-239 count r a t e s were cor rec ted for l o s se s to the

separat ion column and for F . P . contamination - these cor rec t ions will

be t rea ted in section 3.4 - and were normal ized to unit weight of i r ­

radiated UO . The count r a t e was of the order of 25 to 110 c / s e c per

mg of U 0 2 .

3 .4 . Auxil iary measu remen t s

Since the chemical separat ion yields were found to vary some­

what a number of additional measu remen t s were made as a check and

in order to a s s e s s co r rec t ions . The following i tems were checked:

- Np loss to separat ion column, usually of the order of 2 to 3 %.

- Np leakage to F . P . fract ion. A leakage in excess of 0. 5 %

may be detected but was never experienced in these m e a s u r e ­

ments .

- F . P . res idues in the Np f rac t ions . Especial ly Te and Zr

tend to follow the Np.

- The overal l consistency of the chemical separat ion p r o c e s s .

The following measuremen t s were ca r r i ed out.

a) Counting of activity in column filling

The dried glass powder filling was collected and counted in the

single channel analyzer . The count r a t e was added to the count ra te

of the N P sample since neptunium accounted for essent ia l ly al l the

res idua l activity in the filling. The e r r o r introduced through this

correc t ion was at the most about 0.1 % stemming mainly f rom counting

geomet ry differences.

b) Activity sum check

Corresponding to every single i r rad ia ted pellet two or th ree

separat ions were made . The consistency of these separat ions was

checked by adding the various activit ies involved, viz. the count r a t e s

f rom the Np, glass powder, and F . P . samples , and comparing the

r e s u l t s . Thus, also the F . P . samples were counted in the single-

channel analyzer .

- 20 -

In genera l , the agreement between total sums was quite good,

the differences between "ident ical" samples being compatible with the

counting s t a t i s t i c s . Evidently, the separat ion general ly did not in­

volve any var iable non-accountable l o s s e s . The c i rcumstance made

possible an es t imate of the var ia t ion in separat ion yields - Np leakage

to F . P . fraction and/or F . P . contamination in Np fract ion. The

apparent yield variat ion in the Np samples is typically about 1 %.

However, cor rec t ions for variable F . P . res idues (essent ial ly Te-132)

in the Np samples reduced these apparent var iat ions to l e s s than 0. 5 %.

c) Spectrum analyses

The gamma ray spec t ra of a l l samples - N P , F . P . and g lass

powder - were recorded with a Nuclear Data 512-channel analyzer

using a la rge c rys ta l . The purpose was to analyse the spec t ra for the

determinat ion of Np losses or F . P . contamination.

Pu re reference spec t ra of Np-239 and F . P . were p repa red .

In the Np-239 case a 0.1 m m thick natural u ran ium meta l foil was i r ­

radiated under cadmium in the R l la t t ice , thus suppressing strongly

the U-235 fission ra te re la t ive to U-238 capture . The foil was then

processed in the same way as the UO? pellets to give a separa ted Np

fraction with a very slight F . P . contamination. F ig . 6 shows the

spec t rum obtained; the fission product Z r -97 is seen to give a

measurab le contribution even in this case .

A F . P . re ference sample was obtained by separat ing the Np

from a high enrichment meta l piece i r rad ia ted in a t h e r m a l column.

The " p u r e " F . P . spect rum is shown in fig. 7 (the uran ium background

activity is very smal l and is not noticeable). In the corresponding

"Np-fract ion", Np-239 activity was bare ly noticeable, but the par t ia l

and var iable extraction of Te-132 by the separat ion column was

evident.

As mentioned before, Np leakage to the F . P . fraction is be ­

lieved to be l ess than 0. 5 %. This is apparent when comparing the

Np-239 and F . P . spec t ra . Moreover , any leakage is probably con­

stant to within about _+ 0. 2 %, as may be inferred f rom the excellent

uniformity of F . P . spectra in the vicinity of 105 keV.

- 21 -

Examination of the g lass powder activity spec t ra revealed that

the activity was essent ia l ly Np-239 together with one major F . P . ,

namely Te-132 . A smal l cor rec t ion could be es t imated to account for

the la t te r contribution.

The predominant F . P . a r e l is ted below with their main gamma-

ray energies (in keV), compare fig. 7:

T c 9 9 m J±0J^ R u 99 ±4±

j.132 2.3 h ^ x e 1 3 2 231, 670

143 Pr * 294

N b 9 7 T i m i n g M Q 9 7 665> 7 5 Q

Of these the dominant Mo-99 - Tc-99m activit ies and the

smal le r Ce-143 activity appeared to be effectively separa ted ( less than

about 5 % of their total act ivi t ies r emain with the Np). On the con t ra ry ,

Te-132 and Zr -97 were par t ia l ly extracted together with Np in the

separat ion p r o c e s s . The Zr -97 contamination was ra the r smal l and

essent ia l ly constant and was not co r rec ted for; Te-132 , however, con­

tr ibuted a var iable degree of contamination, ranging f rom near zero to

more than 50 % of total Te-132 activity for which a correc t ion was

neces sa ry . The fractional contamination of Te-132 in the Np-239

samples was determined from the 231 keV peak intensity in the F . P .

spec t rum. The size of the correct ion was established by plotting the

differential Te contamination of pa i r s of identical samples ve r sus the

count ra te difference of the corresponding Np samples . The e r r o r

f rom F . P . contamination is largely sys temat ic and is par t ly cancelled

when activity ra t ios a r e formed; the remaining standard e r r o r con­

tribution the these ra t ios will be about _+ 0.2 %. The uncer ta inty

assigned to the activation of each UO_ pellet as determined from the

average of two counted Np samples is general ly Jh 0.35 % (standard

deviation). This includes uncer ta int ies in i r radia t ion geometry ,

chemical p r o c e s s e s , Np yield and counting, except for e r r o r s s tem­

ming from F . P . contamination and the geometry uncertainty in g lass

powder r e s t activity counting. The la t ter uncer ta int ies a r e more

M o 9 9 _ 6 7 _ h ^

T e 1 3 2 _78_h_^

r 1 4 3 33.4 l u

„ 97 17.0 h^ Zr >

- 22 -

readi ly introduced when activity ra t ios a r e formed, since they a r e l a r g e ­

ly sys temat ic in nature and have been est imated to contribute about

_+ 0. 25 % in these r a t i o s . The total root mean square e r r o r of la t t ice

pellet to t h e r m a l column pellet activity ra t ios will then be typically 0. 55 %.

3 . 5 . Cu foil activity counting

The Cu foils were gamma counted in a double-detector s ingle-

channel analyzer with an automatic foil changer . The double-detector

a r rangement served to reduce foil positioning e r r o r s . The Nal(Tl)

c rys ta l s were 1. 75 inch in d iameter by 2 inches high. The g a m m a - r a y

spec t ra of the two detectors were matched using the Cs-137 600 keV l ine.

The pulses f rom the de tec tors were added before amplification and were

counted in a single sca l e r . The analyzer was used as a d i sc r imina to r ,

the threshold energy being in the "valley" below the 511 keV annihilation

peak.

The Co foils were counted in severa l cyc les , thus reducing the

effects of electronic instabil i ty. The drift was checked by counting a

Cs-137 standard sample and was found to be quite smal l ; in fact , no

significant e r r o r from this source has been attr ibuted to the final Cu-64

activity values .

The count r a t e s were automatically punched on paper tape for

subsequent computer t r ea tment . The final Cu activity data obtained

were cor rec ted for background, dead- t ime , decay and foil weight dif­

fe rences . The overal l uncertainty was l ess than jf 0.2 % standard de­

viation in individual foil ac t iv i t ies ; when four foil activit ies were

averaged to give the interpolated activity at the position of a UO_ pellet

this uncertainty was reduced only slightly due to the smal l uncertainty

in the neutron flux distr ibution along the fuel assembly . Cu-activi ty

ra t ios then were assigned an e r r o r amounting to near _+ 0. 3 %.

- 23 -

4. RESULTS

4 . 1 . Activation data

Tables 2 and 3 show the re la t ive Np-239 and Cu-64 activity d i s ­

tr ibutions respect ive ly in the fuel c lus te r . The act ivi t ies a r e nor ­

malized to the average c lus ter activity (N and C respect ively) as ca l ­

culated from

Ä = ~ (A1 + 6A2 + 6A4 + 6A5)

As is expected, the activation distr ibut ions a r e ra ther more flat in the

hot cases (II and III) than in the cold (I). However, the two hot ca ses do

differ somewhat. F r o m table 2 it is c lear that the chemical separat ion

and coincidence methods compare very favourably. The coincidence

data a re those given by H. Pekarek [ 9 ] .

In table 3 the Cu activi t ies a r e compared in cases I and II with

the data obtained by Johansson and Sund [10] in the very same i r r a ­

diations but for near mid-posi t ion of the fuel tes t sect ion. Since the

data agree very well , it may be concluded that at l eas t the rad ia l

t he rma l neutron flux distr ibution is unperturbed by the spacer gap. In

this connection it may also be r e m a r k e d that the effect of the spacer

gap the rmal flux peaking apparently dies away within about 13 cm

along the c lus ter from the gap [10] as determined from the fission

product activity in uranium foils .

In table 4 the Np-239 and Cu activity dis tr ibut ions a r e compared .

It is seen that the distr ibut ions a r e remarkab ly s imi l a r .

Table 2. Relative total Np-23 9 activity in fuel c lus ter normal ized to

the average activity in cluster (N./N).

Fuel pin pos. No.

1

2

5

4

I (35°C)

0.810 + 0.003

0.865 + 0.003

1.054 + 0.004

1.113 + 0.003

I I (212.5°C)

0.835 + 0.003

0.884 + 0.003

1.055 + 0.004

1.088 + 0.004

I I I (212°C)

ohem.sep.

0.842 + 0.003

0.890 + 0.003

1.051 + 0.004

1.086 + 0.004

coino.count

0.842 + 0.003

0.891 + 0.003

1.050 + 0.003

1.086 + 0.003

- 24 -

Table 3 . Relative Cu activation in fuel c lus ter normal ized to the

average activity in the cluster (C . /C) .

Fuel pin pos.

No.

1

2

5

4

I (35°C)

This work

0.774 + 0.002

0.853 + 0.002

1.068 + 0.002

1.116 + 0.002

Johansson <S Sund

0.773 + 0.002

0.849 + 0.002

1.070 + 0.003

1.119 + 0.004

I I (212.5°C)

This work

0.820 + 0.002

0.884 + 0.002

1.056 + 0.002

1.090 + 0.002

Johansson <S Sund

0.818 + 0.002

0.885 + 0.002

1.056 + 0.003

1.089 + 0.003

I I I (212°C)

This work

0.825 + 0.002

0.886 + 0.002

1.054 + 0.002

1.089 + 0.002

Table 4. Comparison between the re la t ive Np-239 and Cu activity

distr ibut ions in a fuel c luster ( N . / N ) / ( C . / C ); the e r r o r

is about ± 0. 55 %.

Fuel pin pos.

No.

1

2

5

4

I (35°C)

1.047

1.014

0.987

0.997

I I (212.5°C)

1.018

1.000

0.999

0.998

I I I (212°C)

1.021

1.006

0.997

0.997

The l a t t i ce - to - the rma l column activity r a t i o s , N./N and C. /C*,

were then formed. These r a t i o s , of cou r se , depend on the actual fluxes

in the Ågesta core and the the rmal column. In case III the resu l t s

yielded systematical ly higher activity ra t ios for Np (by about 1 %) than

the coincidence technique. The discrepancy apparently a rose in the

counting of the the rmal column sample ; probably it was the resu l t of

the compounded e r r o r in the two methods . An unfortunate 1 % uncer ­

tainty for the the rma l column sample was , moreove r , attached to the

coincidence method in this very case . For the chemical separat ion

method the ra t ios N./N' e were associated with an e r r o r of 0.5 to 0.6 %.

- 25 -

The copper foil activity r a t i o s , C . /C* , had an es t imated e r r o r

of about + 0.3 %.

Table 5, finally, gives the values of R? = ( N J / N ^ J A C . / C * ) . The

c lus ter average values have been calculated according to

R1

N ;

N .

It is seen that the two hot cases (II and III) differ by nea r ly 2 % , a d i s ­

crepancy that a r i s e s when the t he rma l column activi t ies a r e taken into

account. This will be discussed further in section 4 .2 in connection

with the p?o r e s u l t s .

Table 5. R? = (N./N*)/(C./C*)

Fuel pin poa. No.

1

2

5

4

Cluster ave.

I (35°C)

1.380 + 0.008

1.337 + 0.008

1.300 + 0.008

1.314 + 0.008

1.320 + 0.007

I I (212.5°C)

1.382 + 0.009

1.357 + 0.008

1.356 + 0.008

1.355 + 0.008

1.357 + 0.007

I I I (212°C)

1.411 + 0.009

1.389 + 0.009

1.378 + 0.009

1.378 + 0.009

1.383 + 0.008

4 . 2 . Derivation of pOQ

Eq. (12) in section 2 may be wri t ten as

p 2 8 + 1 = R = K R '

where R' is a quantity derived f rom two activity ra t ios according to

section 4 . 1 , see table 5, and K is a correc t ion factor:

K = K K 3 /K 2

- 26 -

with

K. = 1 + C / C , (resonance absorption in lat t ice Cu foils)

K_ = 1 + C*/C*(resonance absorption in t he rma l column Cu foils)

K- = 1 + N * / N f + N*/Nf (resonance and fast absorption in t h e r m a l 6 r t s t c o l u m n u»238).

The cor rec t ion factor K. is obtained from the expressen

(Cu is assumed to have a l / v t h e r m a l c ross - sec t ion)

K 1 = l + P R I (13)

where

<*„J TT T /4T + 01» o\J o' n

a = 4 .50 + 0.15 b is the 2200 m / s c ross - sec t ion of Cu [ 1 1 1 . o — '

RI = 1. 92 _+ 0. 20 b is the resonance integral (excluding l / v ) for a 0.10 m m thick foil [12] .

I» = 2 [ E ( k T ) / E I 1 ' 2 a = 0. 90 a with E = 5 E (kT) is the epi-^ ^ the rmal l / v in tegra l .

|3 = r / ( l - 1.01 r) where r is the Westcott epi thermal index.

The quantity r as well as the neutron t empera tu re T have been de te r ­

mined by Johansson and Sund [10] for each of the pins in the fuel

c luster and a r e l isted in table 6 together with the factor K . ; the lat ter

is of the order of 1. 02 to 1. 04 with an e r r o r of about 0. 35 %, about

half of which is due to the e r r o r in the Cu resonance integral .

The other two correc t ion fac tors , K? and K_, per ta in to the

the rmal column act iv i t ies . K_ contains a t e r m to account for fast

neutron absorption, N*/N*, that cannot be neglected although it is

r a ther much smal ler than in a complete c lus te r . Eq. (5) has been

used to es t imate this r a t io . Ref. [13] gives the f a s t - to - the rma l fission

ra t io , F 2 8 / F 2 5 , as (1.81 + 0.08) • 10 for a single UOz rod (corrected

for fission source neutrons other than f rom the rod itself). In our case

the the rmal column sample is only 5. 5 cm long and should give a some­

what smal le r r a t io ; however, the quoted value will be retained as a

- 27 -

f i r s t o rder approximation. Using <x?o - 0.107 and S,. (25)/2 .(28) = 1.52

one obtains

N*/N* = 0. 0029 + 0. 0005

The remaining t e r m s in K_ and K~ c o r r e c t for resonance a b s o r p ­

tion in Cu-63 and U-238; the last t e r m in eq. (13) may be used for the

calculation. In order to es t imate these activation contr ibut ions, the gold

foil cadmium ra t io has been determined in the the rmal column of Rl in

order to a s s e s s 3« The r e su l t s a re given in the following table :

R c d ( T . C . ) - 1 R Cd R d ( l a t t . ) - 1 P

Rl lat t ice (central channel) 3.24 - 0.0420 + 0.0015

Rl the rmal column (T. C. ) at 10 cm from inne uran ium source

-4 10 cm from inner end without 640 285 1.5 • 10

Rl T . C . with 15 cm long , U 0 2 rod {<f> 1. 7 cm) with 120 53 7. 9 • 10 inner end at 10 cm; along rod

Rl T . C . with 5.5 cm long . UOz rod - - 4 . 2 • \VT

The R l lat t ice 3 value has been obtained in other measu remen t s

[14] and the the rmal column 3 values have then been calculated f rom

the cadmium r a t i o s . Using only a 5. 5 cm long UO? rod, the epi thermal

flux component will be reduced accordingly, as shown.

The g determined in this manner is not wholly appropr ia te when

assess ing the Cu-63 and U-238 resonance act ivat ions , however . The

the rma l column sample is a localized source yielding a slowing-down

spec t rum that dec reases slower than l / E when E i n c r e a s e s . There fore ,

abso rbe r s with their main resonance absorption at energies above that

of gold will experience an effectively higher 3 value. In spite of this

effect, the 3 obtained above has been used since the resul t ing e r r o r will

be quite smal l .

- 28 -

The (3 value derived above is not consistent with ep i thermal cap­

ture r a t e s inferred from exper iments by H. Pekarek [ 9 ] . The resul t ing

d i sc repancy is of the order of 2 to 3 % in R or ICR. A further check of

the ep i thermal contribution in a t he rma l column with a fission source is

under way.

The ep i the rma l - to - the rma l activation ra t ios for the Rl column

a re

C*/C* = 2 • 10" 4 and N j / N * = (3.1 + 0.8) . 10" 3

so that

K2 = 1.0002 + 0.0001

K3 = 1 + 0. 0031 + 0. 0029 = 1. 0060 + 0. 0009

K 3 / K 2 = 1.0058 + 0.0010

In the TRIGA the rmal column approximately the same |3 values

as in the case of Rl should obtain, since the epi thermal contribution

f rom the sample fission source dominates over that f rom the r e a c t o r .

The cor rec t ion factor K is reproduced in table 6, and the values

of p ? R = R - 1 a re given in table 7. The difference of 2 % in R com­

paring the two hot cases (II and III) resu l t s in a difference of about 6 %

in p?o» A discrepancy of this order could hardly be caused by dif­

ferent ial F . P . contamination. The discrepancy is probably a com­

bination of random e r r o r s in the the rmal column samples . The effect

of epi thermal flux depletion induced by the spacer gap in case II is

believed to be negligible, as was the effect of the the rmal flux peaking

effect mentioned before . The influence of about 0.8 MWd/kgU burnup

in the adjacent fuel assembl ies should also be unnoticeable. Taking

the mean of the cluster averages of p_ s for cases II and III, one ob­

tains p 2 8 = 0.422 + 0.015.

- 29 -

Table 6. Evaluation of the cor rec t ion factor K

(K-/K = 1. 0058 + 0. 0010 for all cases)

Temp.

(°0

35

212

Pin

pos.

1

2

5

4

1

2

5

4

T n

<°C)

127

118

96

94

405

381

339

333

r

0.065

0.059

0.048

0.046

0.053

0.050

0.044

0.043

K . 1 + C /C 1 r t

1.0362 + 0.0043

1.0325 + 0.0039

1.0258 + 0.0031

1.0246 + 0.0030

1.0378 + 0.0045

1.0352 + 0.0042

1.0294 + 0.0035

1.0293 + 0.0035

K K

K . 1 3

K2

1.0422 + 0.0046

1.0385 + 0.0042

1.0317 + 0.0035

1.0305 + 0.0034

1.0438 + 0.0046

1.0412 + 0.0043

1 .0354+0.0036

1.0353 + 0.0036

Table 7. o2„ = R - 1

Fuel pin pos. No.

1

2

5

4

Ave.

I (35°C)

0.438 + 0.011

0.388 + 0.010

0.341 + 0.010

0.354 + 0.010

0.365 + 0.009

II (212.5°C)

0.443 + 0.012

0.413 + 0.010

0 .404+0 .010

0.403 + 0.010

0.409 + 0.009

I I I (212°C)

0.473 + 0.012

0.446 + 0.011

0 .427+0 .010

0.427 + 0.010

0,435 + 0.010

4. 3. Calculation of p and ICR

The resonance escape probability as expressed in t e r m s of p?„

by eq. (6) with k ,,. = 1 was then de termined.

The c ros s - sec t i ons and p a r a m e t e r s requi red for calculating

(1 - p) a r e given in table 1 0 at the end of this sect ion. The fast dif-2

fusion a r ea L required in determining the corresponding non-leakage

probability P was calculated for an effective resonance neutron energy

- 30 -

of 100 eV using the BURNUP code formula [ 6 ] , A factor of 2 unce r ­

tainty in the energy value cor responds to only a 0.1 % e r r o r in (1 - p). 2

Fo r the buckling B the actual core geometr ica l value was taken.

Table 8 gives the values of p obtained. In all cases the contribution of

the las t t e r m in the equation for (l - p), i . e . the cor rec t ion for fast

neutron absorption a l ready included in the fast fission factor , was

0.0023 + 0.0002.

Table 8. Comparison of measu red and calculated p values

Measurement

I (35°C)

II (212. 5°C)

III (212°C)

Ave. II & III

1 - p (experiment)

0.0977 + 0.0025

0.1142 + 0.0026

0 .1217+0.0029

0.1180 + 0.0040

P (experiment)

0.902 + 0.003

0.886 + 0.003

0.878 + 0.003

0.882 + 0.004

p (theory)

0.892

0.879

0.879

0.879

p - p

exp 'th (*)

+ 1.2

+ 0.8

- 0 . 1

+ 0.4

The e r r o r s quoted originate f rom the uncertainty in p_~ only.

The theore t ica l values of p a re taken from ref. [15] and a r e calculated

with the BURNUP code that uses the resonance integral data of Hell-

s t rand [ 1 6 ] . The theory gives slightly lower values than the exper i ­

mental r e s u l t s . The t empera tu re dependence tends to be somewhat

l a rge r in the experimental c a s e .

The initial conversion ra te was calculated according to eq. (7)

and is shown in table 9. Since the F . P . activity was not measured

accura te ly , no RCR* values have been obtained direct ly but may be

calculated according to eq. (10) by multiplying the ICR by

riiiih 1 + *iat. ^äT^Uh' 1 + «th

The e r r o r in ICR and RCR''C corresponds to that in R = 1 + p->o» since

the c ros s - sec t ion uncer ta int ies have not been included. The theo re ­

t ical ICR has been calculated according to the BURNUP formal i sm [ 1 7 ] ,

- 31 -

T a b l e 9« C o n v e r s i o n r a t i o r e s u l t s

Measurement

I (35°C)

II (212. 5°C)

III (212°C)

A v e . II & III

ICR

(experiment)

0 . 7 7 5 + 0 . 0 0 5

0 . 8 2 5 + 0. 006

0 . 8 4 1 + 0 . 0 0 6

0 . 8 3 3 + 0 . 0 0 8

Relative ICR

1.000

1 .065

1 .085

1 .075

R C R *

1 .392 + 0 . 0 1 0

1 .490 + 0 . 0 1 0

1 .517 + 0 . 0 1 1

1 .504 + 0 . 0 1 5

ICR (theory)

Temp. (°C)

20

220

0 . 7 9 5

0 . 8 3 8

T a b l e 1 0 . L a t t i c e p a r a m e t e r s u s e d in c a l c u l a t i n g o and ICR

1 Quantity

2 2 L ( c m ), s lowing down a r e a S t o 100 eV

2 - 2 B (m ) , g e o m e t r i c a l b u c k ­

l ing of c o r e

P = 1/(1 + L 2 B 2 ) s ' x s '

11 [ 1 5 ]

e [ 1 5 ]

V25

<*28

F / F ( l a t t i c e c l u s t e r [ 1 3 ] S L

a a t ( 2 8 ) (b)

a a t ( U ) (b)

a a t ( 2 5 ) (b)

^ a t ( 2 8 ) / S a t ( U )

S a t < 2 8 > ^ a < 2 5 >

^ ^ s W ^ + ^ s W c o l .

35°C

8 4 . 7

2 . 1 6

0 .9820 + 0 . 0 0 1 0

1 .3115

1 .0268

2 . 4 3

0 .107

0 .052 + 0 . 0 0 1 5

2 . 6 9 8

7 .450

4 . 7 5 2

0 . 3 6 2 1

0 .5678

1 .0067

212°C

1 1 4 . 0

2 . 1 6

0. 9760 + 0 . 0 0 1 5

1 .2928

1 .0276

2 . 4 3

0 .107

0 .053 + 0 .0015

2 . 7 1 0

7 .336

4 . 6 2 6

0 . 3 6 9 4

0 . 5 8 5 8

1 .0120

- 32 -

5 . SUMMARY AND DISCUSSION

The ep i the rma l - to - the rma l neutron capture ra te ra t io (p?o) i*1

U-238 in the Ågesta r eac to r fuel has been measu red using a chemical

method to separa te Np-23 9 activity (being proport ional to U-238 capture)

from uran ium and fission products (F . P . ) . In chemical work of this

type, activity losses a r e liable to occur; neve r the le s s , such lo s ses have

been reduced to an acceptable level . Rather , the fai lure of the chemical

p roce s se s chosen to yield a complete separat ion sets the l imit on ob­

tainable accuracy . A number of auxil iary measu remen t s were requi red

to determine activity losses and res idues in order to a s s e s s co r rec t ions .

These c i rcumstances rendered the measuremen t s and the data analyses

somewhat cumbersome . It might also be argued that sys temat ic e r r o r s

may easi ly a r i s e . However, a comparison of Np-239 activity ra t ios

with r e su l t s from the coincidence method (measuring Np-239 in liquid

samples) has not revealed any large differences between the two

methods .

The incompleteness of the chemical separat ion showed up in the

res idual Np in the separat ion column (2 to 3 %) after separat ion and in

the F . P . , notably te l lur ium and zirconium, remaining with the Np (up

to 2 % of total activity). Np "leakage" to the F . P . fraction was l ess

than 0. 5 % and probably var ied l ess than 0. 2 %. These data a r e the

resu l t s of studies of g a m m a - r a y spect ra obtained with a large detector

c rys ta l .

A ra ther small c rys ta l was used to m e a s u r e the Np-239 activity.

Thus, the potential of reducing the F . P . contamination by means of

pulse height discr iminat ion has not been fully exploited. A further r e ­

lative reduction of the contamination level by 50 % should be possible

using a large c rys ta l . More detailed investigations of the F . P r e ­

sidues should be made by studying the g a m m a - r a y spec t ra with a high-

resolution l i thium-drifted germanium c rys ta l . Any systemat ic e r r o r s

arising from incomplete separat ion yields would in that case be sub­

stantially reduced.

In order to separa te the epi thermal capture r a t e s from the

the rmal the " l / v subtraction technique" was employed. The activation

of copper foils was used to monitor the l / v capture . A complementary

the rmal column i r radiat ion is necessa ry with this method.

- 33 -

In the measu remen t s p ? 8 was de termined ra the r than the so

called modified re la t ive conversion ra t io (RCR*). Either quantity may

be used to calculate p and ICR. Whereas p is obtained more d i rec t ly

f rom p_o than f rom RCR*, the converse is t rue for ICR. The difference

in the determinat ion of p ? f i and RCR*l ies mainly in the attainable p r e ­

cision with which the l / v capture in U-238 is monitored in the one case

(p?o) and the U-235 fission ra te in the other (RCR*). The cor re la t ion

of the two neutron absorption p roces se s - and therefore of p ? f i and

RCR* - does not involve too g rea t uncer ta in t i es . In the fo rmer c a s e ,

the l / v monitor (Cu) is never perfect and a cor rec t ion for resonance

activation has to be applied. When monitoring U-235 fission ra te by

means of F . P . counting, a cor rec t ion for U-238 fission ra te is n e c e s ­

sa ry . However, F . P . counting has one specific advantage, viz . the

favourable i r radia t ion geomet ry . No comparison of the two methods

of monitoring has been made but is cer ta inly des i rab le .

As a conclusion it may be stated that the counting of Np-239-

induced activity by the chemical separat ion method appears to give

reasonably accura te r e s u l t s . The resonance escape probabili ty and

the initial conversion ra t io as calculated f rom p ?„ agree fair ly well

with the theore t ica l values obtained with the BURNUP code employing

resonance integral data .

ACKNOWLEDGEMENTS

The author wishes to thank S. - E . Kroon for his excellent and

skilful work in carrying out the neptunium separat ions and the develop­

ment work that preceeded it . The a u t h o r ' s grat i tude is also due to

Mr . C . - E . Wikdahl and Mr . H. Pekarek for cooperation and helpful

d i scuss ions .

- 34 -

REFERENCES

RYDELL, N. , BLOMBERG, P . E. , and ERICSSON, E. , Experience from the commissioning, the cr i t ica l i ty exper iments and the power operation of the Åge sta Nuclear Power Plant . U.N. Internat . conf. on the peaceful uses of atomic energy, Geneva, 3 , 1964. P r o c . Vol. 5. New York 1965, p. 421 .

APELQVIST, G. et a l . , Reactor physics studies and compar isons between r eac to r physics data f rom calculations and mock-up studies and f rom measuremen t s in the Åge sta Nuclear Power Plant . U.N. Internat . conf. on the peaceful uses of atomic energy, Geneva, 3 , 1964. P r o c . Vol. 5. New York 1965, p. 458.

McHUGH, B . (ed) , The Åge sta Nuclear Power Station. A staff repor t by AB Atom­energ i . Stockholm 1964.

THOMASEN, J. and WINDSOR, H. H. , 2 3 g Measurement of resonance absorption in U using a chemical separat ion technique for isolation of Np^39. 1963. (KR-44).

EGIAZAROV, V . B . , DIKAREV, V . S . and MADEEV, V . G . , Measuring the resonance absorption of neutrons in a u ran ium-graphite la t t ice . Conf. Acad. Sci. USSR on the peaceful uses of atomic energy, July, 1955 (AEC-t r -2435, P . 1 p. 59-68).

AHLSTRÖM, P . - E . , 1961. AB Atomenergi , Sweden (internal repor t RFR-151) .

KOUTS, H. and SHER, R. Exper imenta l studies of slightly enriched uran ium, wate r -moderated la t t i ces . P . l . 1957. (BNL-486).

KROON, S . - E . , 1965.

AB Atomenergi , Sweden (internal repor t AP-RKP-42) (in Swedish).

PEKAREK, H. , AB Atomenergi . Pr iva te communication.

JOHANSSON, H. and SUND, L. , Fine s t ruc ture measurement s in the Ågesta Reactor R3 . 1964. AB Atomenergi , Sweden (internal r epor t F F X - 3 4 1 ; R3-354). HUGHES, P . J . and SCHWARTZ, R . E . , Neutron c r o s s - s e c t i o n s . 2 ed. 1958. (BNL-325).

BERNANDER, G. , Unpublished.

- 35 -

NYLUND, O . , Measurements of the fast fission factor (e) in UO_-e lements . 1961. (AE-40).

JOHANSSON, E. , LAMPA, E. and SJÖSTRAND, N . G . , A fast chopper and its use in the measu remen t of neutron spectr Arkiv för Fys ik _18 (I960) 513-531,(AB Atomenergi , Sweden Internal repor t RFX-43).

APELQVIST, G. , Lattice p a r a m e t e r s for the Ågesta r eac to r fuel. 1962. AB Atomenergi , Sweden (Internal repor t R3-305; RFR-171) (in Swedish).

HELLSTRAND, E. , Measurements of the effective resonance integral in uran ium meta l and oxide in different geomet r i e s . J . Appl. Phys . .28 (1957) 1493-1502.

APELQVIST, G. , State Power Board . Pr iva te communication.

- 36 -

LIST OF FIGURES

1 The Åge sta r eac to r latt ice

2 Axial Cu activation distr ibution along fuel assembly

3 Photograph of exper imental fuel assembly

4 Exper imental fuel assembly . I r radia t ion geometry

5 The rma l column sample

6 Np-239 pulse height spec t rum

7 F i ss ion product spec t rum

8 Np-239 counting geometry

9 Spectrum of unseparated sample

Fig, 1 and 2

• Fuel assembly O Control rod po­

sition

Fig. 1. The Ägesta reactor lat t ice.

Lattice posit ions are identi f ied q u a d r a n t - w i s e using quad rant nota t ion (A toD) and coord inates (without sign).

c

\-,0,

C g > o Ö

o 5-

l Position of test sample Fuel assembly: CIO

Temperature: 212.5 °C

Fig.2. Axial Cu activation distr ibut ion.

200 250 300 31 Axial posit ion-cm f rom bottom plate

Report RFT-142 .

J

a)

Fig . 3 . Ägesta r eac to r fuel assembly a) Top section of o rd inary a s sembly with

coolant tube cut open

b) Special exper imenta l assembly showing demountable set of six fuel pins from an in termediate section.

Fig. 4 and 5,

/Removable set of six 'experimental fuel pins. 17.0

u foils

Experimental ^ p e l l e t

Cu foils

Ordinary pellets

Fig. A. Experimental fuel assembly.

a) Cross section. b) Arrangement of U02 pel lets and Cu foils. The foils were shielded

against fission products by Al or Fe foils.

(A l l measures in mil l imetres.)

5 5.0

Pellet lengths 22.7 20 45 2.0 22.7

^ — É ' 17.0 ^ 0

Experimental pel let.

f> Direction towards reactor.

Cu foi ls.

Fig. 5. Thermal co lumn sample. The sample was clad in Zircaloy tubing.

(Al l measures in mi l l imetres.)

2-105-

0) c c o

SL

o

S.105-JO "c zs o o

V - e s c a p e peak 7 8

F ig . 6. Np 2 3 9 pu lse he igh t s p e c t r u m .

N a l ( T l ) c r y s t a l : Diam. 3 i n . , height 1,5 i n .

Np 2 3 9 ( s e p a r a t e d sample) - re fe rence spec t rum

Np 2 3 9 + U + F. P. ( u n s e p a r a t e d )

Energy values refer to N p " only.

Count ing geometry approx . 40 %

506 F. P. contaminat ion ( Z r 9 7 )

— 1 1-

250 50 100 150 Energy (Channel no.)

200

Fig. 7, Fission product spectrum. (reference)

M o " - ^ T c 9 9 m

K1 kev

Na l (T l ) crysta l ; Diam. 3 in . , height 1,5 in.

Counting geometry approx. 40 %

Z r 9 ^ N b 9 7

665

-r— 50 100 150 200 250

Energy (Channel no.)

Fig. 8 and 3

W/////////A Teflon container

• Np solution

Piexiglas holder

Nal(Tl) crysta l

(1,75 in. x 2 in.)

Photomultiplier tube

(EMI 6097 B)

Fig. 8. .239 counting geometry.

105 keV

Pulse height (volts)

Fig. 9 . Spectrum of unseparated sample with the counting arrangement of f ig . 9

LIST OF PUBLISHED AE-REPORTS

1-220. (See the back cover earlier reports.)

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

222. Total cross-sections of U, UO: and ThOi for thermal and subthermal neutrons. By S. F. Beshai. 1988. 14 p. Sw. cr. 8 : - .

223. Neutron scattering in hydrogenous moderators, studied by the time de­pendent reaction rate method. By L. G. Larsson, E. Möller and S. N. Purohit. 1986. 26 p. Sw. cr. 8 : - .

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

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

228. Theoretical time dependent thermal neutron spectra and reaction rates in H:0 and DiO. By S. N. Purohit. 1966. 62 p. Sw. cr. 8:- .

227. Integral transport theory in one dimensional geometries. By I. Carlvik. 1986. 65 p. Sw. cr. 8:- .

228. Integral parameters of the generalized frequency spectra of moderators. By S. N. Purohit. 1968. 27 p. Sw. cr. 8:- .

229. Reaction rate distributions and ratios in FRO assemblies 1, 2 and 3. By T. L. Andersson. 1965. 50 p. Sw. cr. 8 : - .

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

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

232. Progress report 1965. Nuclear chemistry. Ed. by G. Carleson. 1936. 23 p. Sw. cr. 8 : - .

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

234. Recipient capacity of Tvären, a Baltic Bay. By P.-O. Agnedal and S. O. W. Bergström. 1966. 21 p. Sw. cr. 8 : - .

235. Optimal linear filters for pulse height measurements in the presence of noise. By K. Nygaard. 1966. 16 p. Sw. cr. 8 : - .

236. DETEC, a subprogram for simulation of the fast-neutron detection pro­cess in a hydro-carbonous plastic scintillator. By B. Gustafsson and O. Aspelund. 1966. 26 p. Sw. cr. 8 : - .

237. Microanalys of fluorine contamination and its depth distribution in zircaloy by the use of a charged particle nuclear reaction. By E. Möller and N. Starfelt. 1968. 15 p. Sw. cr. 8:- .

238. Void measurements in the regions of sub-cooled and low-quality boiling. P. 1. By S. Z. Rouhani. 1968. 47 p. Sw. cr. 8 : - .

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

240. Possible odd parity in " sXe. By L. Broman and S. G. Malmskog. 1986. 10 p. Sw. cr. 8 : - .

241. Bum-up determination by high resolution gamma spectrometry: spectra from slightly-irradiated uranium and plutonium between 400—830 keV. By R. S. Forsyth and N. Ronqvist. 1986. 22 p. Sw. cr. 8:- .

242. Half life measurements in 1 "Gd. By S. G. Malmskog. 1966. 10 p. Sw. cr. 8:- .

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

244. Physics experiments at the Agesta power station. By G. Apelqvist, P.-Å. Bliselius, P. E. Blomberg, E. Jonsson and F. Akerhielm. 1968. 30 p. Sw. cr. 8 : - .

245. Intercrystalline stress corrosion cracking of inconel 600 inspection tubes in the Ågesta reactor. By B. Grönwall, L. Ljungberg, W. Hiibner and W. Stuart. 19S6. 28 p. Sw. cr. 8 : - .

246. Operating experience at the Ågesta nuclear power station. By S. Sand­ström. 1986. 113 p. Sw. cr. 8 : - .

247. Neutron-activation analysis of biological material with high radiation levels. By K. Samsahl. 1986. 15 p. Sw. cr. 8:- .

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

249. Optimal linear filters. 2. Pulse time measurements in the presence of noise. By K. Nygaard. 1966. 9 p. Sw. cr. 8:- .

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

251. Absolute transition probabilities from the 453.1 keV level in 183W. By S. G. Malmskog. 1966. 12 p. Sw. cr. 8:- .

252. Nomogram for determining shield thickness for point and line sources of gamma rays. By C. Jönemalm and K. Malén. 1968. 33 p. Sw. cr. 8 : - .

253. Report on the personnel dosimetry at AB Atomenergi during 1965. By K. A. Edwardsson. 1966. 13 p. Sw. cr. 8 : - .

254. Buckling measurements up to 250°C on lattices of Ågesta clusters and on D:0 alone in the pressurized exponential assembly TZ. By R. Persson, A. J. W. Andersson and C-E. Wikdahl. 1966. 56 p. Sw. cr. 8:- .

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

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

257. The Dancoff correction in square and hexagonal lattices. By I. Carlvik. 1966 35 p. Sw. cr. 8:- .

258. Hall effect influence on a highly conducting fluid. By E. A. Witalis. 1966. 13 p. Sw. cr. 8 : - .

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

260. High temperature tensile properties of unirradiated and neutron irradiated 20Cr-35Ni austenitic steel. By R. B. Roy and B. Solly. 1966. 25 p. Sw. cr. 8:- .

261. On the attenuation of neutrons and photos in a duct filled with a helical plug. By E. Aalto and A. Krell. 1966. 24 p. Sw. cr. 8 : - .

262. Design and analysis of the power control system of the fast zero energy reactor FR-0. By N. J. H. Schuch. 1966. 70 p. Sw. cr. 8 : - .

263. Possible deformed states in "Mn and " ' I n . By A. Bäcklin, B. Fogelberg and S. G. Malmskog. 1967. 39 p. Sw. cr. 10:- .

264. Decay of the 16.3 min. 1"Ta isomer. By M. Höjeberg and S. G. Malmskog. 1967. 13 p. Sw. cr. 10:-.

265. Decay properties of 1 "Nd. By A. Bäcklin and S. G. Malmskog. 1967. 15 p. Sw. cr. 10:- .

266. The half life of the 53 keV level in '"Pt. By S. G. Malmskog. 1987. 10 p. Sw. cr. 10:- .

267. Burn-up determination by high resolution gamma spectrometry: Axial and diametral scanning experiments. By R. S. Forsyth, W. H. Blackladder and N. Ronqvist. 1967. 18 p. Sw. cr. 10:- .

268. On the properties of the s , 1, >. a3u transition in 1"Au. By A. Bäcklin and S. G. Malmskog. 1987. 23 p. Sw. cr. 10:-.

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

270. An optical model study of neutrons elastically scattered by iron, nickel, cobalt, copper, and indium in the energy region 1.5 to 7.0 MeV. By B. Hoimqvist and T. Wiedling. 1967. 20 p. Sw. cr. 10:-.

271. Improvement of reactor fuel element heat trans'er by surface roughness. By B. Kjellström and A. E. Larsson. 1937. 94 p. Sw. cr. 10:-.

272. Burn-up determination by high resolution gamma spectrometry: Fission pro­duct migration studies. By R. S. Forsyth, W. H. Blackadder and N. Ron­qvist. 1967. 19 p. Sw. cr. 10:- .

273. Monoenergetic critical parameters and decay constants for small spheres and thin slabs. By I. Carlvik. 24 p. Sw. cr. 10:-.

274. Scattering of neutrons by an anharmonic crystal. By T. Högberg, L. Bohlin and I. Ebbsjö. 1967. 38 p. Sw. cr. 10:- .

275. T h e l A K I = 1 , E1 transitions in odd-A isotopes of Tb and Eu. By S. G. Malm-skog, A. Marelius and S. Wahlbom. 1987. 24 p. Sw. cr. 10:-.

276. A burnout correlation for flow of boiling water in vertical rod bundles. By Kurt M. Becker. 1937. 102 p. Sw. cr. 10:- .

277. Epithermal and thermal spectrum indices in heavy water lattices. By E. K. Sokolowski and A. Jonsson. 1967. 44 p. Sw. cr. 10:-.

278. On the d5 2<-"^97i2 transitions in odd mass Pm nuclei. By A. Bäcklin and S. G. Malmskog. 1967. 14 p. Sw. cr. 10:- .

279. Calculations of neutron flux distributions by means of integral transport methods. By I. Carlvik. 1967. 94 p. Sw. cr. 10:- .

280. On the magnetic properties of the K = 1 rotational band in "»Re. By S. G. Malmskog and M. Höjeberg. 1967. 18 p. Sw. cr. 10:- .

281. Collision probabilities for finite cylinders and cuboids. By I. Carlvik. 1967. 28 p. Sw. cr. 10:- .

282. Polarized elastic fast-neutron scattering of " C in the lower MeV-range. I. Experimental part. By O. Aspelund. 1987. 50 p. Sw. cr. 10:- .

283. Progress report 1966. Nuclear chemistry. 1967. 26 p. Sw. cr. 10:- . 284. Finite-geometry and polarized multiple-scattering corrections of experi­

mental fast-neutron polarization data by means of Monte Carlo methods. By O. Aspelund and B. Gustafsson. 1987. 60 p. Sw. cr. 10:-.

285. Power disturbances close to hydrodynamic instability in natural circulation two-phase flow. By R. P. Mathisen and O. Eklind. 1987. 34 p. Sw. cr. 10:- .

288. Calculation of steam volume fraction in subcooled boiling. By S. Z. Rou­hani. 1967. 26. p. Sw. cr. 10:-.

287. Absolute E1, A K = O transition rates in odd-mass Pm and Eu-isotopes. By S. G. Malmskog. 1967. 33 p. Sw. cr. 10:-.

288. Irradiation effects in Fortiweld steel containing different boron isotopes. By M. Grounes. 1967. 21 p. Sw. cr. 10: .

289. Measurements of the reactivity properties of the Ågesta nuclear power reactor at zero power. By G. Bernander. 1967. 43 p. Sw. cr. 10:- .

290. Determination of mercury in aqueous samples by means of neutron activa­tion analysis with an account of flux disturbances. By D. Brune and K. Jir-low. 1967. 15 p. Sw. cr. 10:- .

291. Separtaion of 5'Cr by means of the Szilard-Chalmers effect from potassium chromate irradiated at low temperature. By D. Brune. 1967. 15 p. Sw. cr. 10:-.

292. Total and differential efficiencies for a circular detector viewing a circu­lar radiator of finite thickness. By A. Lauber and B. Tollander. 1987. 45 p. Sw. cr. 10:- .

293. Absolute M1 and E2 transition probabilities in " 5 U . By S. G. Malmskog and M. Höjeberg. 1967. 37 p. Sw. cr. 10:- .

294. Cerenkov detectors for fission product monitoring in reactor coolant water. By O. Strindehag. 1967. 58 p. Sw. cr. 10:-.

295. RPC calculations for K-forbidden transitions in 1e3W. Evidence for large inerfial parameter connected with high-lying rotational bands. By S. G. Malmskog and S. Wahlbom. 1967. 25 p. Sw. cr. 10:-.

296. An investigation of trace elements in marine and lacustrine deposits by means of a neutron activation method. By O. Landström, K. Samsahl and C-G. Wenner. 1967. 20 p. Sw. cr. 10:-.

297. Natural circulation with boiling. By R. P. Mathisen. 1967. 58 p. Sw. cr. 10:- . 298. Irradiation effects at 160—240°C in some Swedish pressure vessel steels.

By M. Grounes, H. P. Myers and N-E. Hannerz. 1967. 38 p. Sw. cr. 10:- . 299. The measurement of epithermal-to-thermal U-238 neutron capture rate (p 2 8 )

in Ågesta power reactor fuel. By G. Bernander. 1967. 42 p. Sw. cr. 10:—.

Förteckning över publicerade AES-rapporter

1. Analys medelst gamma-spektrometri. Av D. Brune. 1961. 10 s. Kr 6:—. 2. Bestrålningsförändringar och neutronatmosfär i reaktortrycktankar — några

synpunkter. Av M. Grounes. 1962. 33 s. Kr 6:- . 3. Studium av sträckgränsen i mjukt stål. Av G. Ostberg och R. Attermo

1963. 17 s. Kr 6:- . 4. Teknisk upphandling inom reaktorområdet. Av Erik Jonson. 1963. 64 s.

Kr 8:- . 5. Ågesta Kraftvärmeverk. Sammanställning av tekniska data, beskrivningar

m. m. för reaktordelen. Av B. Lilliehöök. 1964. 336 s. Kr 15:-. 6. Atomdagen 1965. Sammanställning av föredrag och diskussioner. Av S.

Sandström. 1966. 321 s. Kr 15:-. Additional copies avaiable at the library of AB Atomenergi, Studsvik, Ny­köping, Sweden. Micronegatives of the reports are obtainable through Film-produkter, Gamla landsvägen 4, Ektorp, Sweden.

EOS-tryckerierna, Stockholm 1967