AECL-5964

ATOMIC ENERGY G f f & L'ÉNERGIE ATOMIQUEOF CANADA LIMITED E f i j T DU CANADA LIMITÉE

BURNUP DETERMINATION OF NUCLEAR FUELSUSING NEODYMIUM-148

by

T. T. Vandergraaf, L. AA. Carefoot and D. G. Boase

Whiteshell Nuclear Research EstablishmentPinawa, Manitoba ROE 1LO

June, 1978

ATOMIC ENERGY OF CANADA LIMITED

BURNUP DETERMINATION OF NUCLEAR FUELS USING NEODYMIUM-148

by

T.T. Vandergraaf, L.M. Carefoot and D.G. Boase

Whiteshell Nuclear Research Establishment

Pinawa, Manitoba, ROE 1L0

June, 1978

AECL-5964

DETERMINATION DE LA COMBUSTION DU COMBUSTIBLENUCLEAIRE PAR L'UTILISATION DU NEODYME-148

par

T.T. Vandergraaf, L.M. Carefoot et D.G. Boase

RESUME

Cette méthode décrit la façon de déterminer la combustion des

combustibles d'uranium et de thorium-uranium en mesurant la concentration148

de Nd. Cette méthode implique la separation du neodyme du combusti-

ble irradié dissous au moyen d'une combinaison d'extraction liquide -

liquide et de chromatographie par échange d'ions. On détermine ensuite

le Nd par spectrométrie de masse de :

sant du Nd comme diluant isotopique.

le Nd par spectrométrie de masse de la dilution isotopique en utili-

142Nd

L'exposition de l'analyste aux rayonnements est considérable-

ment réduite en enlevant le Ce/Pr, le Cs et le Cs au début de

la séparation. La précision de la méthode est meilleure que 4% au

niveau 2a et la différence entre cette méthode et la détermination par235

appauvrissement de U est de 3%.

Ce rapport décrit le développement de la méthode de séparation

du néodyme et comprend des méthodes de séparation chimique et d'analyses

détaillées ainsi que les calculs requis pour déterminer la combustion.

L'Energie Atomique du Canada LimitéeEtablissement de Recherches Nucléaires de Whiteshell

Pinawa. Manitoba ROE 1L0Juin, 1978

AECL-5964

BURNUP DETERMINATION OF NUCLEAR FUELS USING NEODYMIUM-148

by

T.T. Vandergraaf, L.M. Carefoot and D.G. Boase

ABSTRACT

A method is described for determining the burnup of uranium148

and thorium-uranium fuels by measuring the Nd concentration in the

fuel. The procedure involves separation of neodymium from dissolved,

irradiated fuel by a combined liquid-liquid extraction/ion-exchange

chromatography procedure followed by determination of Nd by isotope142

dilution-mass spectrometry, using Nd as isotopic diluent. Radiation144

exposure to the analyst is reduced considerably by removing Ce/Pri34 237

Cs and Cs in the early stages of the separation procedure. Theprecision of the method is better than 4% at the 2a level and the bias

235between this method and burnup determination by U depletion is 3%.

This report describes the development of the neodymium separa-

tion procedure and includes detailed chemical separation and analysis

procedures and the calculations required to determine the burnup.

Atomic Energy of Canada LimitedWhiteshell Nuclear Research Establishment

Pinawa, Manitoba ROE 1L0June, 1978

AECL-5964

CONTENTS

Page

1. INTRODUCTION 1

2. BURNUP BY THE l40Nd METHOD 7

3. DETERMINATION OF NEODYMIUM 9

3.1 SEPARATION FROM IRRADIATED FUEL 93.1.1 Development of the Separation Procedure 93.1.2 Performance Data 17

3.2 SEPARATION PROCEDURE 183.3 MASS SPECTROMETRIC ANALYSIS 183.4 CALCULATIONS 20

4. DETERMINATION OF URANIUM 20

5. DETERMINATION OF THORIUM 21

6. BURNUP CALCULATIONS 22

7. DISCUSSION OF THE METHOD 27

7.1 ACCURACY 277.2 COMPARISON WITH OTHER METHODS 317.3 PRECISION 32

7.4 ASSESSMENT OF THE METHOD 33

8. SUMMARY 34

9. ACKNOWLEDGEMENTS 34

10. REFERENCES 35

APPENDICES 37

1. INTRODUCTION

Burnup in nuclear fuels may be defined as the energy produced

by a given amount of fuel. It is commonly expressed in terms of MW.h/kg

HE or GJ/kg HE, that is, the number of megawatt hours or gigajoules

produced in one kilogram of heavy element isotopes (thorium, uranium,

plutonium). Burnup is a fundamental measurement of nuclear fuel to

which phenomena such as grain growth, swelling, fission gas release, and

general fuel behaviour can be related.

For low enrichment uranium fuels, the most widely used tech-

nique for burnup determination is based on measurement of the number of

U atoms consumed in the irradition. This number is determined by235 238

mass spectrometric measurement of the U/ U ratio in samples of the

original fuel and of the irradiated fuel. The total number of fissions235 (1)

of U can then be calculated from a knowledge of the neutron cross-235

sections for fission, a,, and for capture, fJp, in U. Since the ratio

of these cross-section values is virtually constant for all operating235

conditions of a specific nuclear reactor, the number of fissions in U235

can be related to the decrease in the number of U atoms by the equation:

4 V [<N35>0 - <Vt]where: Nf = number of fissions per unit weight of fuel,

235(N.q)n = number of U atoms at time t = 0, per unit weight•" ° of fuel,

235(N__). = number of U atoms at time t, per unit weight of" C fuel

235 238or, expressed as U/ U ratios:

- 2 -

N°f

f ac + af cN38

where: I o.n I and 1 0 1 Q I are the isotopic ratios before and after

the irradiation interval respectively, and N 3 g is the number of U

atoms per unit weight of fuel. This method works very well for enriched235

fuels with an initial U concentration of 2 to 5% of the total uranium

for a bui-uup of 700 to 1100 GJ/kg U (200 to 300 MW.h/kg U), or for

natural fuel with a burnup of approximately 300 to 600 GJ/kg U (80 to

150 MW.h/kg U). In other cases, the method is not suitable for the

following reasons:

1. In high enrichment, low burnup fuels, the change in the235 238

U/ U ratio is small and consequently, the error associated

with the measurement of the isotope ratios is large in relation

to the difference between these ratios. This is reflected in

a large error of N,, and hence, burnup.

2. In low enrichment, high burnup fuels, the contribution to the239

total number of fissions by Pu becomes appreciable as shown239

by Figure 1. This number of Pu fissions must be determined

to obtain the correct burnup value. Normally, this is calcu-

lated using the appropriate cross-sections, decay constants

and branching ratios of the actinides involved and is subject

to significant error.

3. In mixed fuels of uranium-thorium, uranium-plutonium and233 239

uranium-thorium-plutonium, fissions of U and Pu may

account for a major fraction of the total number of fissions235

and the U depletion method is invalid.

I

10

I

200 400 600Burnup GJ/kg U

800 1000

9-3Q lit

FIGURE 1 J Pu AND U FISSION FRACTIONS VERSUS BURNUP FOR PICKERING FUEL

- 4 -

To illustrate, the changes in the various uranium isotopic,233.. .235., 234,,,233TT 236,. .235.. . 235,. .238.,. .

ratios ( U/ U, U/ U, U/ U, and U/ U; in

uranium-thorium fuels with burnup are shown in Figure 2 for235

two different initial U/Th ratios. The higher the thorium233

content, the more U is produced and thus the isotope

ratios change. In addition, uncertainties in the nuclear

constants will be reflected in the accuracy and precision of

the burnup values calculated solely on the basis of actinides.

A second group of techniques used to determine burnup is based

on the formation of fission products. Every fission event produces two

fission products in addition to a fixed amount of energy. If the con-

centration of a particular fission product can je determined, the number

of fissions per unit weight of fuel can be calculated and from this, the

burnup. Ideally, to be used as a burnup indicator, a fission product

should:

1. have the same fission yield for all fissioning isotopes of

interest in the fuel under investigation,

2. not be present in the fuel as an impurity,

3. not be formed during irradiation by any other process, for

example: by activation of an impurity in the fuel, by acti-

vation of other fission products, or by decay of long-lived

fission products,

4. not be removed from the matrix during irradiation, that is, it

should have a small absorption cross-section for neutron

capture, have precursors with short half-life and/or small

absorption cross-sections, have no gaseous precursors, be

stable or have a long half-life and not migrate through or be

lost from the fuel, so that its distribution is different from

the fission distribution.

- 5 -

1.0

0.9

i

EXPERIMENTAL ThO2/UO2FUEL

0 = 2.79xlO17 n.m^.s

-H- = 0.0266 (93.5% U)Th

^7 = 0.0293 (93.5%Th

I I

25

20

15

100 200 300 400 500 600 700 800 900 1000 1100 1200

Burnup (GJ/kg HE)

FIGUEE 2 EFFECT OF INITIAL U/Th RATIO ON URANIUM ISOTOPICRATIOS AS A FUNCTION OF BURNUP

- 6 -

Further, it should

5. be easily separated chemically from the fuel and

6. be readily determined by an accurate analysis procedure.

No isotope meets all these requirements, but for uranium and

uranium-plutonium fuels, Nd meets most of them. In addition, one142

neodymium isotope, Nd, is not produced in fission since its precursor,142 142

Ce, is stable. Any Nd present in irradiated fuel must therefore

have been introduced into the fuel as an impurity before irradiation ,

and its concentration can be used to determine the amount of Nd from

sources other than the fission process.

This is not completely correct, since some Nd can be produced235

from two sources, namely in fission of U - via the stable fission141

product Pr:

235... .. .... . . 141Pr(n,Y)1A2Pr B" 142NdU(n,f) mass 141 chain " l g - .

141and from naturally occurring Pr, which may be present in the fuel

as an impurity:

141Pr (n,y) W 2 P r19.2 h

The first reaction involves two neutron capture processes, and pro-

duces an insignificant amount of Nd which can be ignored for all

practical purposes.

The 141Pr/142Nd ratio in the earth's crust is * 0 . 8 ^ . If this

abundance ratio is maintained in the fuel, the increase in the

in the fuel by neutron capture in Pr will amount to less than

at a burnup of 1400 GJ/kg HE.

U2Nd

- 7 -

2. BURNUP BY THE '*°Nd METHOD

The procedure for determining the burnup of uranium, uranium-

thorium, uranium-thorium-plutonium, and uranium-plutonium fuels requires

determination of the concentrations of Nd, total uranium, and where

applicable, total thorium and plutonium, in an aliquot of the irradiated

fuel solution. The number of fissions that have occurred in a given

quantity of the fuel may then be calculated and the burnup in GJ/kg HE

obtained from a knowledge of the energy produced per fission.

The main difficulty in using Nd to determine burnup arises

from the similarity of its chemical behaviour to that of other rare

earths formed and the consequent complexity of the separation procedure

required to isolate neodymium of sufficient purity to permit its accurate

determination. For uranium-thorium fuels, a further difficulty arises

due to the difference in Nd fission yields from the two main fissioning

isotopes, U and U (1.28 and 1.69% respectively). As the irradiation

of a mixed uranium-thorium fuel proceeds, the fraction of the total number233 148

of fissions due to U increases, leading to a decrease in the Nd

yield and a corresponding uncertainty in the overall fission yield for the

irradiation. This problem can be resolved by calculating an effective

fission yield for Nd based on the fraction of the total fissions233

which occur in U:

Y* = xY3 + (l-x)Y5

where: Y = effective fission yield for Nd,

Y 3 = fission yield for 148Nd from 233U (1.28%),

Y5 = fission yield for 148Nd from 2 3 5U (1.69%),

fraction of the fissions occurring in U.

This fission fraction is readily obtained from reactor physics calcula-(3>

tions using the computer code LATREP •

- 8 -

Experimental ThO,/UO, fuels irradiated in the WR-1 reactor235

normally contain 1 to 10% UO,, enriched to 93.5% U, and 90 to 99%

ThO9. Data generated by LATREP show that in fuel containing 10% UO™235 233

(93.5% U) only 2% of all fissions occur in U up to a burnup of1440 GJ/kg HE (4TJ MW.h/kg HE) and that an overall error in burnup of

235less than 0.5% arises if it is assumed that all fissions occur in U.

233 235However, for fuels containing 1.3% IK>2 the U/ U fission ratio at

1440 GJ/kg HE is 1:1 at this burnup level and a weighted fission yield

must be used in the burnup calculations. In general, the fission ratio

in a particular fuel can be calculated using LATREP with an uncertainty

of 7 to 10%, which corresponds to an uncertainty in the burnup calcula-

tion of 1 to 1.5%.

Table 1 summarizes the errors in burnup calculated by this

approach which accrue from the uncertainty in the fission yield.

TABLE 1

BURNUP ERROR AS A FUNCTION OF UNCERTAINTY IN233U/235U FISSION RATIO

BURNUP INDICATOR: 1/>

BURNUP = 1440 GJ/kg HE (400 MW.h/kg HE)

FUEL COMPOSITION BURNUP ERROR CONDITIONS

Th0o

90 %

98.7%

98.7%

(93%

10 % 0.48%

1.5 %

2.8 %

Assuming all fissions due to

Assuming 10% uncertainty in

" J U / u fission ratio

Assuming 20% uncertainty in233U/235U f ± s s l o n r a t i o

- 9 -

3. DETERMINATION OF NEODYMIUM

3.1 SEPARATION FROM IRRADIATED FUEL

3.1.1 Development of the Separation Procedure

A number of procedures for the separation of neodymium from

irradiated fuel have been reported^ " . These methods usually consist

of a preliminary separation of uranium and plutonium from the fission

products and trans-plutonium elements, followed by separation of the

rare earth elements*as a group from other fission products and finally

the isolation of neodymium from the rare earths. The final step is

difficult because of the similarity in chemical behaviour of the lantha-

nides. Neodymium is usually determined by isotope dilution mass spectro-

metry, although neutron activation analysis has also been used '

Our initial experiments employed an ion-exchange method and

were directed at adapting the two-column separation procedure developed

by Abernathey et al at Los Alamos . A flow diagram of this method is

shown in Figure 3. The low neodymium recoveries (̂ 35%) obtained were

traced, at least in part, to the adsorption of neodymium on glassware,

especially during the final ion-exchange step. By using polyethylene

and polypropylene labware, the neodymium yields were increased to > 70%.

Neodymium fractions of sufficient purity were difficult to

obtain. Trace amounts of rare earth elements, especially praseodymium

and promethium, interfered in the mass spectrometrlc determination,134 137 144 144

while small amounts of Cs, Cs, Ce, and Pr in the final

solution taken for mass spectrometry gave radiation levels of approxi-

mately 0.1 mGy/s @ 1 cm (100 mR/h @ 0.1") which are higher than desirable

for a mass spectrometric determination.

- 10 -

15O.T, 233.. 242DNd, U, Pu

TRACERS

FUEL SOLUTION

- 6 mol.L"1 HNO3

EVAPORATE TO DRYNESS

DISSOLVE IN 12 mol.L"1 HC1

AG 1 x 10

50-100 mesh

ELUTE WITH 12 mol.L"1 HC1/0.1 mol.L"1 HI

ELUTE WITH 0.1 mol-L"1 HC1 (U

12 mol.L"1 HC1

EVAPORATE TO DRYNESSDISSOLVE IN 0.8 mol.L"1 HNO,/90% MeOH

AG 1 x 2200-400mesh

ELUTE WITH 6 mL 0.08 mol.L"1 HNO_/90% MeOH

ELUTE WITH 6 mL 0.08 mol- L"1 HNO3/9O% MeOH

0.8 mol.L"1 HNO3/90% MeOH

FIGURE 3 FLOWCHART OF THE SEPARATION PROCEDURE FOR NEODYMIUMFROM IRRADIATED FUELSOURCE: ABERNATHEY et al IAEA-SM 149/37 VIENNA 1972

- 11 -

A disadvantage of the Los Alamos method is that cerium accompa-

nies neodymium through the procedure and is retained on the final ion-

exchange column. Cerium-144 has a half-life of 284 days and decays via144 144

17.3-minute Pr to stable Nd by By emission. The energies associ-

ated with the g decay are 2 to 3 MeV. Since plastic labware is used

throughout this procedure and since only small amounts of liquid are

used, (3 shielding is minimal and there exists a considerable radiation

hazard to the analyst. This can be overcome by using shielded facili-

ties, and/or remote handling but this complicates the procedures and

does not get at the root of the problem, that is, the continued presence

of large amounts of radionuclides throughout the procedure.

An improved procedure was developed, based on the fact that

cerium can be oxidized to Ce(IV) with potassium bromate in acidic solu-

tions and can then be extracted into di-(2-ethylhexyl) phosphoric acid

(HDEHP)^ . Uranium, thorium, plutonium, and zirconium also readily

extract into HDEHP and this step can therefore be used instead of the

first column in the previous procedure. The rare earths, transplutonium

elements (americium and curium), radiocesium and large amounts of

potassium and bromine remain in the aqueous phase. However, potassium

and bromine are only sparingly soluble in the methanolic nitric acid

loading and eluting solutions used in the final ion-exchange column and

therefore interfere in this separation step.

(12)Zelenay has reported a separation procedure for neodymium

from irradiated fuels in which the rare earths are isolated from bromine

and potassium by reverse phase chromatography using HDEHP as the sta-

tionary phase and 0.1 mol.L HC1 as the eluent. The rare earths,

americium and curium are extracted into the organic phase and back-

extracted into 0.3 mol.L"1 HC1.

This procedure was adapted by us for liquid-liquid extraction.

The rare earths are back-extracted into 1 mol.L UNO, rather than into

- 12 -

0.3 mol.L HC1, since the former is more compatible with the methanolic

nitric acid solution used to load the partially separated neodymium on

the ion-exchange column.

(13)While this method was being developed, Marsh et al reported

that a number of analysts had encountered difficulties in applying the

final ion-exchange step used to isolate neodymium from other rare earths.

This was traced to differences in the elution characteristics of various

batches of the AG 1-X10 ion-exchange resin used. Also, their experience

showed that an appreciable amount of neodymium remained adsorbed on the

resin. For these reasons, a macroporous resin, AGMP-1 , was recommended

and the compositions of the loading and eluting solutions were changed

from 0.78 mol.lT1 HNO,/90% MeOH and 0.0078 mol.L"1 HN0,/90% MeOH to 1.56— 1 — 1 Cl"i}

mol.L - HN03/80% MeOH and 0.094 mol.L HNOj/80% MeOH respectivelyv J;.

The performance of the latter ion-exchange resin was tested

using synthetic mixtures of praseodymium, neodymium and samarium and

analyzing the eluent by emission spectroscopy. Some results are shown

in Figure 4, and indicate that an adequate separation can be obtained.

The reproducibility of the neodymium elution profile was determined

using Nd tracer solutions and counting 2 mL fractions of the eluent

by y spectrometry. Results for four runs are shown in Figure 5.

3 —1The low flow rates through the ion-exchange columns (0.1 cm .min

(1.7 mm .s )) dictated the use of a metering device to deliver the**

eluting solution to the columns at the correct rate. A Technicon pump

proved unsuitable since minute amounts of plasticizer were leached from

the tubing and affected the quality of the mass spectrometric determina-

tion of the neodymium isotopic ratio. Accordingly, a syringe pump was

designed and built (see Figures 6 and 7) to overcome this problem. It

has a capacity of four 30-mL syringes and delivers the eluting solution

at a rate of 0.093 ± 0.002 cm .min" (1.55 mm ,s~ ).

* Bio Rad Laboratories (Canada Ltd.), Hississauga, Ontario

** Technicon International of Canada Limited, Montreal, P.Q.

AGMP-1 ANION EXCHANGE COLUMN200-400 mesh

ELUENT: 0.094 mol .L" 1 HNO3/80% MeOH

6 12 15 18

ELUENT VOLUME (mL)

21 24 27 30

FIGURE 4 ELUTION PROFILE OF SELECTED RARE EARTH ELEMENTS FROM A IIACROPOROUS COLUMN

60

50

CO . - ^

S 40

S 3 0

20

10

I I \AGMP-l ANION EXCHANGE COLUMN

200-400 mesh

T i i rELUENT: 0.094 mol-L-i HNO3/80% MeOH

-RUN #3

RUN #1

10 12 14 16 18 20 22 24ELUTION VOLUME (mL)

26 28

FIGURE 5 ELUTION PROFILE FOR NEODYMIUM FROM A MACROPOROUS COLUMN

- 15 -

IiVO

W

OM

I

FIGURE 7 SYRINGE PUMP AND ION EXCHANGE COLUMNS USED IN FINAL NEODYMIUM SEPARATION

- 17 -

3.1.2 Performance Data

3.1.2.1 Decontamination Factors

Decontamination factors, defined as:

„ _ concentration of element in original solutiond concentration of element in final solution

are given in Table 2.

TABLE 2

DECONTAMINATION FACTORS FOR SELECTED RADIONUCLIDES

Element K,d

Cs 600*

Ce > 300***

Pr -v 100**

Sm ^ 10

* Average of 10 runs

** Average of 2 runs

3.1.2.2 Nd Recovery

1. Solvent Extraction Step 80%**

2. Ion-Exchange Step 65%

3. Overall Recovery 55%

* Average of 10 runs

** Average of 4 runs

- 18 -

3.2 SEPARATION PROCEDURE

A schematic of the complete separation procedure is shown in

figure 8. Detailed operating instructions are given in Appendix G.

142A weighed aliquot of a Nd tracer solution is added to a

weighed aliquot of the fuel solution to be analyzed. The acidity is

adjusted to ̂ 10 mol.L HNO., and solid potassium bromate (KBrO_) is

added to oxidize cerium(III) to cerium(IV). Uranium, plutonium, thorium,

cerium and some of the zirconium and ruthenium are extracted with HDEHP,

previously conditioned with KBrO». The aqueous phase is evaporated to

dryness, converted to the chloride form with concentrated HC1, dissolved

in 0.1 mol.L~ HC1 and extracted with HDEHP. The rare earths and the

higher actinides (americium and curium) are extracted into the organic

phase, leaving radiocesium, bromine and potassium in the aqueous phase.

Americium, curium, and the rare earths are- then back-extracted from

HDEHP into 1 mol.L~ HNO-, evaporated to dryness and redissolved in a—1

1.56 mol.L HNO,/80% methanol loading solution. This solution is

passed through a previously prepared ion-exchange column containing 200

to 400 mesh AGMP-1 macroporous resin. Americium, curium, and the rare

earths of higher atomic number than neodymium (Pm to Lu) are eluted in

12 mL of 0.094 mol.L~ HN03/80% methanol solution, and the neodymium is

eluted in the next 9 mL. This neodymium fraction is concentrated by

evaporation to •*< 100 uL.

A second portion of the sample solution, not containing Nd

tracer, is carried through the identical separation procedure and both

samples are analyzed by mass spectrometry for isotopic composition.

3.3 MASS SPECTROMETRIC ANALYSIS

The concentrated neodymium fractions are loaded by evaporation

on resistance heated tantalum filaments. The filaments are mounted in

- 19 -

EVAPORATE TO DRYNESSCONVERT TO CHLORIDE FORMDISSOLVE IN O.I mol.L"1 HC1

0000Q0(DISCARD)

AQUEOUS

©0©(DISCARD)

EVAPORATE TO DRYNESSDISSOLVE TN1.56 mol -L"1 HNOj/80% MeOH

(DISCARD)

AGMP-1200-400mesh

0.094

0.094 mol.L* /80% MeOH 13-22 mL_ [ Nd

FIGURE 8 FLOWCHART FOR THE CHEMICAL SEPARATION OF NEODYMIUMFROM IRRADIATED FUEL

- 20 -

the triple filament source of a CEC Model 703A thermal ionization mass

spectrometer. The 140 to 150 m/e range is scanned cyclically at least

15 times, the signals digitized and the 148Nd/142Nd isotopic ratio cal-

culated by computer. Detailed procedures described by Shields are

followed in general.

3.4 CALCULATIONS

Detailed calculations to obtain the Nd concentration in the

fuel aliquot are given in Appendix H.

4. DETERMINATION OF URANIUM

Since the burnup is calculated from the Nd concentration in

the fuel, the concentration of the major heavy elements, uranium and

thorium (in thorium-uranium fuels), in the fuel solution must be deter-

mined .

For uranium fuels, a modified Davies-Gray procedure ' is

normally used in our laboratories. It is precise (< 1% at 2a) and

virtually interference-free ' but requires more than 10 mg of

uranium and, since with irradiated fuels this is accompanied by a high

level of radiation, the use of a shielded facility is necessary.

An alternative, less precise method uses the spectrophotometric(19)

uranium peroxide procedure with measurement of the absorbance at 400 nm

The precision of a single determination is "v 4% at 2a and the technique

requires as little as 50 vg of uranium.

Consolidated Electrodynamics Corporation, Monrovia, California

- 21 -

A more advantageous method for uranium determination is isotope235

dilution mass spectrometry using U as the tracer for the low enriched

or depleted fuels, and U for highly enriched fuels. It is precise (y

1% at 2a) and can be performed on submicrogram samples, treliminary

separations of uranium from fission products are carried out in a

shielded facility, reducing the radiation exposure to the analyst.

For uranium-thorium fuels, isotope dilution mass spectrometry

is the preferred method for uranium determination since the ratios of

uranium to fission products are too low to use titrimetric or spectro-

photometric methods in unshielded areas.

Detailed analysis procedures are described in Appendices D and

E.

5. DETERMINATION OF THORIUM

Thorium is determined by complexometric titration using

ethylenediaminetetracetic acid (EDTA) at a pH of 2.9 with xylenol orange

indicator. The precision is <\< 4% (2a) at a level of 4 to 5 mg. Elements

that interfere with the titration are usually absent from fuel solutions.

Some zirconium fuel sheath is dissolved with the fuel but is not present

in large enough quantities to affect the accuracy of the thorium determi-

nation.

No alternative procedures to determine thorium with the nec-

essary precision have been developed. Isotope dilution mass spectromety230

using Th is feasible, but has not yet been investigated, owing to the

high cost of this tracer.

Detailed analysis procedures are described in Appendix F.

- 22 -

6. BURNUP CALCULATIONS

The burnup equation using Nd requires the following:

1. Nd concentration in the fuel aliquot,

2. U, Th, Pu concentrations in the fuel aliquot, where applicable,

3. yield for Nd from the various fissioning nuclides,

4. fission energy of the various fissioning nuclides.

3 and 4 can be combined to give a quantity k, defined as:

k = number of Nd atoms created during the production

of 1 Joule of energy. (1)

For 2 3 5U, the fission energy is 3.22X1011 Joules (200.8 MeV)i/o 215

and the Nd fission yield from D is 1.69%. Thus, 100 fissions in

U produce 1.69 Nd atoms and release 3.22x10 Joules of energy,

and therefore:

k = 1.69/3.22xl0"9 = 5.25xlO8 148Nd atoms/Joule.

Table 3 gives some k values applicable to mixed oxide fuels:

- 23 -

Fissioning

Isotope

2 3 3U

2 3 5U

239Pu

241Pu

TABLE 3

k VALUES FOR VARIOUS FISSIONING NUCLIDES

Fission Yield148Nd (X) (20)

1.28

1.69

1.69

1.91

Fission Energy

(MeV) ( 2 )

197.5

200.8

209.3

212.5

k atoms Nd/J

4.05xl08

3.25xlO8

5.04xl08

5.61xlO8

where:

The burnup equation can be expressed as follows:

Nft n aBU (GJ/kg HE) = Tj8- 2 T-i

F i=l Ki

148Nd8

WF

atoms of in aliquot of fuel,

(2)

weight of heavy elements in fuel aliquot,

fraction of total number of fissions occurring

in ith fissioning isotope,

atoms Nd per Joule.

The number of Nd atoms is related to the number of Nd

tracer atoms by the equation:

N8 " N2 (l-(3)

- 24 -

where: N~, Ng = number of Nd and Nd atoms, respectively,

M = 148Nd/142Nd ratio in the fuel/spike mixture,

T = 148Nd/142Nd ratio in the 142Nd spike solution,

S = 148Nd/142Nd ratio in the fuel solution,

or, expressed in concentrations:

[N8] =

where: [N_],[NO] = Nd and Nd concentrations in spike and

fuel solution respectively,

a = weight of spike or tracer solution,

b = weight of fuel solution.

Combining (2) and (4)

OU b{[U]+[Th]+[Pu]} (5)

where: W_ in (2) is expressed here as a product of fuel aliquot withr

weight (b) and of heavy element concentration in gram per gram

of fuel solution.

For uranium and uranium-thorium fuels, the weight of plutonium

can be ignored, and the first term of (5) reduces to:

2 (6)b{[U]+[Th]}

235In cases where U accounts for essentially all fissions, such as in

enriched or natural uranium fuels, the equation (5) can be simplified

to:

- 25 -

BU =5.25x10

11[-T \M/S J(M-T

1-M/S(7)

Example: Natural U0_ Fuel,

a = 0.6020 g,

1.38xl016 atoms

b = 1.020 g,

[U] = 6.34 mg/g,

M = 0.4623,

S = 63.35,

*T = 0.00058

Nd/g,

BU5.25x1011

0.4623-0.000581-0.4623/63.35

(0.6020)(1.38xl016)(1.020)(6.34)

= 1138 GJ/kg U.

239In cases where Fu fissions must be considered, such as in

Q

natural UO» with high burnup, the value for k of 5.25x10 may be re-

tained; this is an approximation which introduces a maximum error of 2%.

o

For mixed uranium-plutonium fuels, again a value of 5.25x10

for k can be used. In addition, the plutonium content in the fuel must

be known in order to calculate the burnup.

BU =a[N]

b[U]{(Pu+U)/U}5.25x1011

M-T1-M/S

As was discussed earlier, nixed thorium-uranium fuels pose an

additional problem. The fission yields for Nd from U and 2 ^ U

are sufficiently different that the previous approximation in the k

value cannot be used without introducing a substantial error. In these

- 26 -

cases, it is advisable to determine the relative number of fissions in233 235 (3)

U and D by means of the LATREP computer program . In addition,

both the thorium and uranium concentration in the fuel solution must be

determined. A typical example is:

BU

ThO2

a

[N2]

b

[Th]

[U]

M

S

a(233U)

a(235U)

(0.(]

= (8.

/U02 Fuel, 7% U0£

0.

1.

1.

9.

0.

0.

63

0.

0.

,6020)(lL.020)(9

,31xlO14

6020 g,

38xlO16 atom 142Nd/g,

020 g,

2 mg/g,

6 mg/g,

4623,

• 35,

2*.

8.

.38xlO16)

.2+0.6)/ 0.2

U.05X1011

)(2.02xl0~12)(0.4651)

= 780 GJ/kg HE.

233 235If all fissions are assigned to only U or to U, erroneous

burnup values are produced. For example, assuming

1) only 233U fissions: k(233U) = 4.05xl08

This gives a burnup value of 954 GJ/kg HE and an error of

[(954-780)/780]100«22%

215 21"i ft2) Only "V fissions: kC"v) - 5.25x10°

This gives a burnup value of 738 GJ/kg HE and an error of

[780-738/780]100 - -5%

- 27 -

7. DISCUSSION OF THE METHOD

7.1 ACCURACY

The burnup determination based on Nd is considered to be

the most accurate method available and has been accepted by the

American Society for Testing and Materials for uranium and uranium-

plutonium fuels because the Nd/J ratio is virtually independent of

irradiation conditions.

The burnup of uranium-thorium fuels is calculated from the

results of three analyses: determination of thorium by complexometric

titration, uranium determination by redox titrimetry or isotope dilution

mass spectrometry, and determination of Nd by isotope dilution mass

spectrometry.

The titration of uranium is performed with potassium dichromate

standardized against U.0o, an accepted primary standard in the nuclear

industry. None of the fission products in the fuel solution are present

at high enough levels to cause a bias in the measurement . Similarly,

the reagents used in the complexometric titration of thorium are standard-

ized against stoichiometric ThO_ and against pure zinc metal. Interfer-

ences in the titration may be caused by cations that form complexes with

EDTA, zirconium and calcium. Zirconium from dissolved fuel sheath is

not present in sufficiently large quantities to create a bias, and

calcium contamination is avoided by the use of borosilicate glassware

throughout the titration procedure.

142Solutions of the isotopic diluents used in this method, Nd,

235 238U, and U, are standardized chemically against U,0o or NdnCL. Both

JO 2 3of these standards are assayed gravimetrically for uranium and neodymium

- 28 -

content. The only conceivable way a bias can be introduced in the

burnup determination is if Nd from sources other than the fission

process or if isobars are included in the Nd measurement.

The Nd in the fuel solution comes from two sources:

contamination with natural neodymium, either present in the fuel at the

time of manufacture, or introduced as a contaminant during the dissolu-

tion, and from the fission process. The contribution from the natural142

neodymium can be determined from the amount of Nd found in the irradi-142

ated fuel, since essentially no Nd is produced by the fission process

(see above).

148Other rare earth isobars can introduce a bias in the Nd

determination. Both Ce and Sm are naturally occurring isotopes

and both are also formed in the fission process. Their presence in the

neodyraium fraction will affect the measurement of the Nd/ Nd ratio

and for this reason, the mass 140, 147 and 149 peaks are monitored. The

mass 140 peak is monitored to determine the cerium content of the isolated

neodymium fraction. No other isotope of mass 140 has a sufficiently

long half-life to be present li\ irradiated fuel after a decay of a few

months, and therefore the mass 140 peak can only be due to cerium. The

natural Ce/ Ce ratio is 0.13 while the ratio of the fission yields

for the mass 142 and mass 140 chains is 0.92. In a theoretical worst

case, in which all cerium in the fuel is assumed to be produced by the

fission process, the contribution to the mass 142 peak by cerium would

be 0.92 of the mass 140 peak. A correction must then be applied to the

mass 142 peak.

Similarly, the presence of a mass 147 peak indicates that

either Pm or Sm is present. The Sm/ Sm isotopic ratio in

naturally occurring samarium is 1.1 and the fission yield ratio for the

mass 147 and mass 149 chains is 2.1.

- 29 -

A measured mass 147/149 ratio greater than 2.1 indicates the147 147

presence of Pm in the neodymium fraction, and, since Pm is the

only long-lived promethium isotope produced in the fission process, no

correction to the mass 148 peak is required. A mass 147/149 ratio

approaching a value of 2.1 indicates the presence of samarium and that

Sm may impart a bias to the Nd measurement.

Table 4 lists the mass spectrometric results for 14 separated

neodymium fractions together with the intensity of the mass 140, 147 and

149 peaks. In alj. cases, the 147/149 ratio is large, indicating that

little if any samarium is present in the neodymium fraction and the

Sm contribution to the mass peak can be ignored. The 142/140

ratio is high in most cases, and contribution in the mass 142 peak by142

Ce can be ignored. Thus, the bias in the isotope dilution massspectrometric measurement by isobaric contamination is negligible.

The magnitude of the error that can be introduced when Nd

is used to determine the burnup of thorium-uranium fuels is discussed in

the Introduction, and is within the overall error of the measurement of

the Nd concentration in the fuel (Table 1).

- 30 -

TABLE 4

ISOTOPIC MASS SPECTROMETRIC ANALYSES OF NEODYMIUMFRACTIONS SEPARATED FROM IRRADIATED FUEL

Values in weight percent

FUEL SAMPLE

FDO-68001-1

FD0-68001-1T

FDO-68001-2T

EXP-221-WLF

EXP-221-WLF-T

EXP-221-WLH-T

PICK 74-T

PICK 65-T

PICK 50-T

PICK 74

948-1 T

948-2 T

948-3 T

948-4 T

140

0.57

0.17

0.20

0.06

0.07

0.09

0.02

0.05

0.11

0.10

0.02

0.02

0.02

0.34

142

5.83

86.22

86.96

0.15

22.81

26.01

12.80

14.87

12.85

0.29

10.09

6.38

9.13

8.91

147

0.37

0.02

0.02

0.13

0.10

0.19

0.09

0.39

0.59

0.54

0.43

0.87

0.36

0.32

148

8.45

1.14

1.07

10.00

7.79

7.45

8.40

8.17

8.51

9.06

7.65

8.09

7.81

7.71

149

< 0.02

< 0.003

< 0.003

< 0.02

< 0.02

< 0.01

< 0.02

< 0.02

< 0.02

< 0.02

< 0.02

< 0.02

< 0.02

< 0.02

150

4.06

0.57

0.54

3.95

3.16

2.96

4.10

4.08

4.70

4.98

3.10

4.08

3.38

3.34

- 31 -

7.2 COMPARISON WITH OTHER METHODS

235The U depletion method is being used within Atomic Energy

of Canada Limited (AECL) as the standard destructive chemical burnup235 (1)

analysis for natural and slightly enriched (< ̂ 5% U) uranium fuelsTable 5 gives a comparison between burnup values obtained by the Nd

235technique and those obtained by the U depletion method. Agreement is

good with a difference of 3% between the two sets of values. This differ-

ence is probably equal to or less than some of the uncertainties in the

values of the various parameters used in the LATREP computer program to235

calculate burnup values for the U depletion data.

TABLE 5

COMPARISON OF TWO METHODS FOR BURNUP DETERMINATIONON PICKERING U02 FUELS

*Bundle

13900C

13721C

15553C

15527C

15527C

13866C

14672C

Element

11

46

50

65

70

74

76

Burnup (GJ/kg U)

235U Depletion Method

634

702

738

648

655

720

713

148Nd Method

601

738

716

601

626

738

680

Mean

Ratio of BurnupResults

235U/148Nd

1.05

0.95

1.03

1.08

1.05

0.98

1.05

- 1.03 ± 0.09 (2a)

Identifying numbers used within AECL

- 32 -

7.3 PRECISION

Table 6 shows the results obtained from several duplicate

analyses of U0- fuel solutions.

TABLE 6

RESULTS OF REPLICATE BURNOP ANALYSES OF

IRRADIATED UO, FUEL USING 148Nd

Fuel Element Sample Burnup (GJ/kg U)

WN78 1 1 9012 9203 888

x 903 ± 32 (2a)

WN78 5 1 7862 808

x 797

925A CE144 1 8312 870

x 850

920A B/E152 1 4132 415

x 414

928C WAH19 1 2292 228

x 228

Ko replicate analyses data have been produced for mixed thorium-

uranium fuels*

- 33 -

7.4 ASSESSMENT OF THE METHOD

The Nd procedure for burnup determination as developed in

our laboratory offers a marked improvement over those described in the

literature ~ , in that the radiation exposure to the analyst is

reduced by a factor of 2 to 4, without resorting to cumbersome shielding.

The exposure is of course a function of the burnup of fuel and of the

time between irradiation and analysis and consequently, no radiation

dosages per analysis can be quoted.

The ahalysis and separation procedure are lengthy. Total

elapsed time is two days for uranium fuels, and three days for thorium-

uranium fuels, exclusive of mass spectrometer time. Four fuel solutions

can be analyzed and separated together, bringing the total laboratory

time down to 0.5 to 0.75 day per sample. Table 7 gives a breakdown of

the analysis requirements for the various fuels.

TABLE 7

ANALYSIS REQUIREMENTS FOR 1

Step

Chemical Separation

Mass Spectrométry

Titration

Spectrophotometry

Total number ofseparate analyses

Fuel

UU/ThUU/ThUU/ThUU/Th

UU/Th

Number of

Nd U

*8Nd BURNUP

Separations or Analyses

U

IDMS* Titration Spec.

2222

1NR**

U(Titration)

5NR

1NR

U

IDMS*

2222

Th

Titration

1

U(Spectrophotometry) U(IDMS*)

5NR

89

* Isotope Dilution Mass Spectrometry

** Not recommended

- 34 -

8. SUMMARY

Burnup determination of natural uranium fuels based on the235

depletion of U in the fuel during irradiation becomes less accurate

for natural uranium fuels as the burnup approaches 600 GJ/kg U (150

MW.h/kg U) and is not applicable to mixed uranium-thorium fuels. For

this reason, a method based on the Nd production in the fuel has been

developed.

Uranium, thorium (where applicable) and Nd concentrations

in an irradiated fuel solution are determined by isotope dilution mass

spectrometry and titrimetry, after the appropriate separations have been

made. The burnup is a

expressed in GJ/kg HE.

made. The burnup is calculated from the Nd/(U + Th) ratios and

235Agreement with the existing II depletion method is better

than 4%. Precision based on duplicate and triplicate analyses of the

same fuel solution is < 4% at the 2a confidence level.

9. ACKNOWLEDGEMENTS

The authors would like to express their thanks to Mr. C.R.

Hillier for the emission spectrographic analyses, to Mrs. H. Olchowy for

Y spectrometric analyses and to Mr. D. Bell, Mrs. C. Provencal and Mrs.

D. Marek for the mass spectrometric analyses. Mr. R.G. Thomson's con-

tribution in adapting the Los Alamos two-column procedure for use in

this laboratory is also recognized.

- 35 -

10. REFERENCES

1- R.G. Hart, M. Lounsbury, R.W. Jones and M.J-F. Notley,Nucl.-Sci. Eng., 18, 6 (1964).

2. Handbook of Chemistry and Physics, 49th éd., p. F144, TheChemical Rubber Co., Cleveland, 1969.

3. G.J. Phillips and J. Griffiths, "LATREP User's Manual",Atomic Energy of Canada Limited Report, AECL-3857 (1971).

4. Atom Percent Fission in Uranium and Plutonium Fuel (Neodymium-148 Method), ASTM-E321-75, p. 806, American Society for Testingand Materials, Philadelphia, 1975.

5. W.N. Bishop, S.F. Marsh, R.L. Williams and G.E. Wolfe, "Post-irradiation Examination of Thoria-Urania Fuel Rods. IsotopicAnalysis of the Fuel. Final Report", Babcock and Wilcox Co.Report, BAW-3809-7 (1969).

6. B.F. Rider, C.P. Ruiz, J.P. Peterson, Jr., P.S. Luke, Jr.and F.R. Smith, "Accurate Nuclear Fuel Burnup Analysis June1963 - Aug 1963", General Electric Co. Report, GEAP-4361 (1963).

7. L. Koch, G. Cottone and M.W. Geerlings, Radiochim. Acta, 10(3-4),122 (1968).

8. R.M. Abernathey, G.M. Matlack and J.E. Rein, "Sequential IonExchange Separation and Mass Spectrometric Determination ofNeodymium, Uranium, and Plutonium in Mixed Oxide Fuels forBurnup and Isotopic Distribution Measurements", Los AlamosScientific Lab. Report, LA-DC-12810 (1970).

9. H. Ruf, A. v. Baeckmann and E. Gantner, Mikrochimica Acta[Wien], 5, 1029 (1970).

10. M.R. Monsecour and A.C. Demildt, Anal. Chem., .41(1), 27 (1969).

11. J. Krtil, M. Bezdek and J. Mencl, J. Radioanal. Chem., 1/5,»,369 (1968).

12. T. Zelenay, Radiochem., Radioanal. Lett., 2{\), 33 (1969).

13. S.F. Marsh, M.R. Ortiz, R.M. Abernathey and J.E. Rein, "ImprovedTwo-Column Ion Exchange Separation of Plutonium, Uranium, andNeodymium in Mixed Uranium-Plutonium Fuels for Burnup Measure-ment", Los Alamos Scientific Lab. Report, LA-5568 (1974).

- 36 -

14. W.R. Shields, Instrumentation and Procedures for IsotopicAnalysis, U.S. National Bureau of Standards Technical Note277, Washington, D.C. (1966).

15. W. Davies and W. Gray, Talanta, 11, 1203 (1964).

16. R.W. Dyck and D.G. Boase, "Fuel Reprocessing with TertiaryAmine Extractants", Atomic Energy of Canada Limited Report,AECL-4214, p. 22 (1973).

17. "Annual Progress Report for the Period July 1973 through June1974", USAEC New Brunswick Lab., N.J., NBL-272, p. 5 (1974).

18. "Annual Progress Report, July, 1974 - June 1975", EnergyResearch and Development Administration, New Brunswick Lab.,N.J., NBL-277, p. 4 (1976).

19. R.W. Dyck and D.G. Boase, "Fuel Reprocessing with TertiaryAmine Extractants", Atomic Energy of Canada Limited Report,AECL-4214, p. 28 (1973).

20. M.E. Meek and B.F. Rider, "Compilation of Fission ProductYields", General Electric Co., Pleasanton, Calif., NEDO-12154-1 (1974).

- 37 -

APPENDIX A

DISSOLUTION OF IRRADIATED FUEL

A.I Uranium Fuel

A.I.I Reagent-1

7.8 mol.L nitric acidPrepare by diluting analytical reagent grade concentrated nitric acidwith an equal volume of distilled water.

A.1.2 Procedure(To be carried out in a Hot Cell)

1) Cut an ^ 5 mm thick section offuel with sheath attached(<v 10 grams).

2) Dissolve the fuel in <v. 100 mL of7.8 mol.L"1 HNO. under reflux.

3) Transfer about 3 mL of the fuelsolution to a new, stopperedvial and retain for analysis.

A.2 Thorium-Uranium Fuel

•• Notes

Avoid contamination of fuel withother materials in the hot cells.

If cesium assay is required, cutfuel with an argon-cooled sawblade.

Use clean glassware.

Some fission products (Ru, Pd, Mo,Tc) will remain as a black, insol-uble residue which can be ignored,because it does not interfere withthe subsequent analysis. The sheathwill not dissolve.

Discard remainder of the fuelsolution.

A.2.1 Reagent -1 -113 mol.L nitric acid/0.025 mol.L hydrofluoric acid.

Dilute 825 mL of analytical grade concentrated nitric acid to 1 litrewith distilled water in a polyethylene container and add 1 mL concen-trated hydrofluoric acid.CAUTION; Rubber gloves should be worn when handling concentrated

hydrofluoric acid since it is extremely corrosive.

A.2.2 Procedure(To be carried out in a Hot Cell)

1) Cut a <\» 5 mm thick fuel sectionwith the sheath attached (̂ 10 g).

2) Dissolve the fuel with the sheathin <v. 100 mL of 13 mol.L"1 HNO-/0.025 mol.IT1 HF under reflux!

3) Transfer about 20 mL of the fuelsolution to a new, stooperedvial and retain for sampling.

NotesAvoid contamination with other ma-terials, especially natural fuels.

Use clean glassware.Etching of glassware by HF isinsignificant.All fission products will dissolve.

Discard the remainder of the fuelsolution.

- 38 -

APPENDIX B

PREPARATION AND STANDARDIZATION OF TRACER SOLUTIONS

All reagents must be analytical reagent grade.

B.I Neodymium

1. Md tracer solution (y 40 ng/mL)142

Dissolve ^ 2 mg Nd2O3, enriched > 95% in Nd (Union Carbide Corp. OakRidge, TN, USA) in a small amount of ^ 1 mol.L"* HNO3 and dilute to50 mL in a polyethylene volumetric flask with distilled water.

2. Standard natural neodymium stock solution

Heat 99.999% pure Nd2O3 CSpex Industries Inc. Metuchen N.J.) at > 920°Cto constant weight, to remove any carbonates that may be present. Weighout accurately ^ 58 mg M2O3 and dissolve in a small amount of 1 mol.L"1

HNO^ and dilute to 50 mL with distilled water in a weighed polyethylenevolumetric flask. Weigh final solution to obtain net solution weight.

3. Standard natural neodymium solution.

Dilute 5 mL of this solution to 100 mL with distilled water in a poly-ethylene volumetric vial. This dilution should be done on a weightbasis for highest accuracy.

4. Standardization142

Weigh out equal amounts of the Nd tracer solution and the standardnatural neodymium solutions and mix thoroughly.

5. Evaporate portions of the ' Nd tracer solution (step 1), the standardnatural neodymium stock solution (step 2), and the mixed solution (step 4),on tantalum filaments by applying a low voltage current to the filament.

6. Mount the filaments in the thermal ionization mass spectrometer and, usingstandard mass spectrometric techniques .± scan the 140 to 150 m/e rangeat least 15 times. Calculate the I48Nd/142Nd isotopic ratios from thesemeasurements.

7. Calculations142NdThe concentration of the Nd tracer solution is calculated by the

equation:

14 a WNN_ - 9.917xlO14 S2 b U - ( M | / T | ) ]

where: N. - concentration of Nd atoms in tracer solution in atomsper gram

a - weight in grams of natural neodymium standard solution

b * weight in grams of Nd tracer solution

WN - concentration of Nd in natural neodymium standard solutionin wg/g

- 39 -

2 142/144Mr- = Nd ratio in mixed standard-tracer solution

2 142/144Or- = Nd ratio in natural neodyoium standard solution

T f = 1 4 2 / 1 4 4Nd ratio in 142Nd tracer solution

The following is an illustrative example:

let a = 1.0662 g

b = 1.0281 g

WN = 4.711 ug Nd/g

M | = 3.9078

c| = 1.1535

tf| = 120.86

N - o oi7vin14 (1.0662)(4.711)(3.9078-1.1535)2 " »*5-1/x-LU (1.0281) (1-3.9078/120.80)

= 1.379x10 atom Nd per gram solution

B.2 Uranium

Natural Uranium Tracer

)Dissolve "v 10 grams of U3O8 uranium standard (NBS 950a) in a small amountof ̂ 8 mol.L"1 HNO3 and dilute to 100 mL with distilled water in a taredvolumetric flask to give an uranium concentration of ^ 100 rng/g. Weighthe filled flask and record the density of the uranium solution.

2. Enriched Uranium Tracer235

Dissolve *> 10 grams of l^Og, highly enriched in U (NBS U930) in a smallamount of 1» 8 mol.L"1 HNO3 a n d d i l u t e to 10Q mL with distilled water ina tared volumetric flask to give an uranium concentration of <v 100 mg/g.Weigh the filled flask and record the density of the uranium solution.

3. Standardize both solutions by evaporating suitable aliquots in taredplatinum crucibles followed by igniting at 800°C for one hour. Cool ina desiccator and weigh as U,0_.

4. Analyze aliquots of both tracer solutions by mass spectrometry for isotopiccomposition, using the procedure described by Shields^).

5. Dilute the stock solutions by weight as required and calculate the concen-trations in the diluted solutions.

National Bureau of Standards, Washington, D.C., 20234, U.S.A.

- 40 -

APPENDIX C

SAMPLING PROCEDURE

To be carried out in a Hot Cell.

Procedure1. Dilute 1.5 to 2.0 mL of fuel

solution to -v 10 mL with1 mol.L-1 HNO .

2. Transfer 1.5 to 2.0 mL of thisdiluted fuel solution to a cleanscrew-capped glass vial, and re-move this from the hot cell to aventilated fume hood containingadequate shielding.

3. Dilute this fuel solution in aclean, borosilicate glassculture tube to *v 12 mL with 10mol.L"1 HNO- (Solution A).

4. Weigh out an aliquot of solutionA (Wp-) and add accurately weighedout amounts of uranium (Wj) andneodymium tracers (WTN) and mixthoroughly (Solution B).

NotesAn approximate measurement of the fuelsolution is adequate at this stage.

Heavy element concentration should be about15 to 20 mg/mL.

Heavy element concentration will be<v 2 mg/mL.

1} From an estimate of the burnup, claculatethe amount of tracers to be added to give14%d/142Nd and 235U/238U ratios closeto unity.

2) For enriched uranium and thorium-uraniumfuels, use the natural uranium tracer(B.2.1). For natural or depleted uraniumfuels, use the enriched uranium tracer(B.2.2).

- 41 -

APPENDIX D

URANIUM DETERMINATION BY ISOTOPE DILUTION MASS SPECTROMETRY

D.I Reagents

All reagents must be analytical reagent grade.

1. 30% tributylphosphate in carbon tetrachloride (30% TBP/CC1,). Dilute30 mL of TBP to 100 mL with CCI,.

2. 7.8 mol.lT nitric acid.Dilute concentrated nitric acid with at» equal amount of distilled water.

3. 3 mol.L nitric acid.Dilute 100 mL of concentrated HNOo to 500 mL with distilled water.

4. 0.1 mol.L hydroxylamine nitrate in 1 mol.L nitric acid.Dilute 6 mL of a 15% aqueous solution of NH-OH.HNO, to 100 mL with1 mol.L"1 HNO3.

5. 10% ammonium fluoride.Dissolve 10 grams NH.F in 100 mL of distilled water.

D.2 Procedure NotesTo be carried out in a well venti-lated fumehood with adequateshielding.

1. Prepare the following extraction If thorium-uranium fuels are to besolution in 15 mL capped glass analyzed, substitute 1 mL distilledbottles. water for the 1 mL 10% NH4F to pre-5 mL 30% TBP/CC14

V e n t Palpitation of ThF4.

2 mL 7.8 mol.L"1 HNO.1 mL 10% NH4F

2. Transfer 1 mL aliquots of fuel U and Pu are extracted into thesolutions A and B (C.3.4) to organic phase.the capped bottles, shake vig-orously and allow to stand for10 to 15 minutes, for thephases to separate.

3. With a clean Pasteur pipette Wa Lng is done by vigoroustransfer 2 to 4 mL of the or- she :ing.ganic phase to a 15 mL centri-fuge tube and wash with a 1 mLaliquot of 3 mol.L"1 HNO3, cen-trifuge and discard the aqueousphase.

- 42 -

4. Wash the solvent phase 3 times Plutonium is reduced to Pu(III)1 mL aliquots of the NI^OH.HNO- and back-extracted into thesolution»' Centrifuge after aqueous phase.each wash and discard theaqueous phase.

5. Transfer the organic phase to a clean centrifuge tube and back-extract the uranium with two 1 mL aliquots of UJO. Transfer thewashes to a polypropylene beaker.

6. Evaporate to dryness and redissolve the residue in 200 pL of 3mol.L HNO

7. Evaporate the solution on a tantalum filament by applying a lowvoltage current to the filament.

8. Mount the filaments in a thermal ionization mass spectrometer andcyclically scan the 230 to 240 m/e ranges at least 15 times using thetechnique described by Shields'^*'.Calculate the isotopic composition from these measurements.

D.3 Calculations

The uranium concentration in the fuel solution is calculated from theequation:

_î ik- ik 1-1 l k 1

UF UT WF T W_, C..-A., nF l k l k • ï B M

i=l 1 K 1

where: U,, = uranium concentration in the fuel solution in mg/g,r

U_ = uranium concentration in the natural U tracer in mg/g,

W_ = weight of fuel solution (see C.4),r

W™ = weight of tracer solution (see C.4),

A.. = isotopic ratio of uranium isotopes i and k in fuel1 solution,

B., = isotopic ratio of uranium isotopes i and k in uraniumtracer solution,

C.. = isotopic ratio of uranium isotopes i and k in fuel/tracermixture,

M. * mass of i uranium isotope.

The following is an illustrative example:

UT - 87.2 ug/g

W F - 1.069 g

W - 0.502 g

- 43 -

uranium isotopic composition, a'o

233 234 235 236 238 235/238 (m/e)

Fuel 30.30 2.42 41.01 12.25 14.02 2.9271

Tracer - - 0.75 99.25 0.0075Mixture 14.49 1.13 19.96 5.83 58.60 0.3404

v A „ _ 30.30 2.42 41.01±fx

Aik*Mi " 14.02 n3 + 14.02 234 + Ï O 2 235

236 + 238 = 1675.56

238 = 239'78

1675.56/239.78 = 6.99n /nZ A...M. /t B...M, =i=l lk y i=l ik i

IT = «7 9 0-502 /o.0075-0.3404\F 0/<i: 1.069 " ^0.3404-2.9271^ b>

= 36.83 yg U/g fuel solution.

- 44 -

APPENDIX E

URANIUM DETERMINATION BY REDOX TITRIMETRY*

This method is applicable only to uranium fuels. Mixed uranium-thorium fuels re-quire too large a sample to be handled safely outside the hot cell area.

The determination is performed on the diluted fuel solution, A, obtained in C.3,above.

E.I Reagents

All reagents should be analytical reagent grade.

1. 1.5 mol.L sulfamic acid:

Dissolve 150 g of sulfamic acid in 1 litre of cold water.

2. Phosphoric acid/ferrous sulfate solution:Add 50 mL of 1 mol.L"! ferrous sulfate solution to 400 mL concentratedorthophosphoric acid.

3. 8 mol.L" nitric acid/0.15 mol.L" sulfamic acid/0.4% ammoniummolybdate solution:Dissolve 4.0 g of ammonium paramolybdate in 400 mL of water. Add 500 mLof concentrated nitric acid and mix. Add 100 mL of 1.5 mol.L"! sulfamicacid and mix.

4. Dilute sulfuric acid/vanadyl sulfate solution:Add 200 mL of concentrated sulfuric acid to 3 litres of water. Add 3 gof vanadyl sulfate dihydrate. Dissolve and dilute to 4 litres.

5. 0.04% barium diphenylaminesulfonate solution:Dissolve 0.2 g of barium diphenylaminesulfonate in about 200 mL boilingwater. Cool and dilute to 500 mL with water.

6. 0.05N potassium dichrornate:Using primary standard grade potassium dichrornate, accurately weighapproximately 2.45 g. Dissolve in water and make up to 1 litre.

7. Standard uranium solution (100 mg/mL) (to be used to standardize thedichrornate solution):Dissolve 21 g of uranyl nitrate hexahydrate in 25 mL of 1 mol.L'1 HNO3.Dilute to 100 mL in a volumetric flask. Standardize the solution byevaporating a suitable aliquot in a platinum crucible and igniting at850°C for 1 hour. Cool in a desiccator and weigh as Uo0o.

E.2 Procedure

NOTE: This procedure must be carried out in a well-ventilated fume hoodwith adequate shielding against radiation.

The method used here Is that described by W. Davies and W. Gray . It hasbeen modified for use in these laboratories by R.W. Dyck and D.6. Boase(16).

- 45 -

Weigh out an aliquot (W) of thediluted fuel solution A (C.3)containing up to 20 mg uraniuminto a 250 mL conical flask.The total volume of sample mustnot exceed 15 mL.

Add 5 mL of 1.5 mol.Lacid. Mix.

-1sulfamic

Add 45 mL of the phosphoric acid/ferrous sulfate solution, 10 mLof the nitric acid/sulfamic acid/ammonium molybdate solution. Mixthoroughly after each addition.

Wait 2 'to 5 minutes after theappearance of the green colour.

Add 100 mL of the sulfuric acid/vanadyl sulfate solution and 2mL of the barium diphenylamine-sulfonate indicator solution.Mix well after each addition.

Within 10 minutes, titrate to anintense violet colour which doesnot fade for up to 1 minute.

Repeat the above procedurewithout the sample present toobtain a reagent blank.

Calculate the uranium concen-tration by the method givenbelow.

NotesThe aliquot should contain lessthan 3 mmol of nitrate ion.

Sulfamic acid is used to destroynitrate ion which interferes withthe reduction.

The solution will turn dark brown,but will become light green, orcloudy, but pale-coloured, afterabout 30 seconds.

E.3 Calculations

Calculate the uranium concentration of the sample from:

Uranium (mg/g) = 243(T-TB) (B)

where T

TBW

wsample titer (mL of dichromate),

blank titer (mL of dichromate),

sample aliquot (g),

concentration of potassium dichromate (tng/raL),

- 46 -

i

APPENDIX F

THORIUM DETERMINATION BY COMPLEXOMETRIC TITRATIQN

F.I Reagents

All reagents should be analytical reagent grade.

1. 0.1 mol.L* zinc chloride:Weigh out accurately 6.537 g of reagent grade zinc metal and dissolve in<\- 6 mol.L-1 HC1. Dilute to 1000 mL with distilled water.

2. 0.1 mol.L~ ethylenediaminetetracetic acid:Dissolve ^ 37 g Na_ EDTA in distilled water and dilute to 1000 mL. !Standardize against ZnCl_ standard.

3. 0.1 mol.L~ bismuth nitrate:Dissolve ^ 50 g Bi(NO ),.5H 0 in dilute HNO and dilute to 1000 mL withdistilled water. J J J

4. Xylenol orange indicator:Dissolve 0.1 g xylenol orange in 100 mL distilled water.

5. 10 w/v hexamine buffer:Dissolve ^ 10 g hexamine in 100 mL distilled water.

6. Ammonium acetate buffer:Prepare a saturated solution of ammonium acetate in distilled water.

F.2 Standardization

F.2.1 Standardization of EDTA against ZnCl-.

Procedure Notes

1. Dilute 100 mL of ZnCl2 solution to1000 mL with distilled water.

2. Dilute 100 mL of EDTA solution to1000 mL with distilled water.

3. Transfer an accurately weighed This aliquot should contain 0.1 tovolume of 0.01 aol.L"1 ZnCl2 0.5 mmol of Zn.standard to a 100 mL beaker con-taining ^ 50 mL of water.

4. Stir the solution, add 4 drops of The pH should be kept in thisxylenol orange and adjust the pH range during the titration.to 5.5 to 6 by adding hexaminebuffer solution.

- 4 7 -

5. Titrate to a lemon yellow endpointusing 0.01 mol.L-1 EDTA.

6. Repeat titration without the ZnCl,. To obtain a reagent blank.

Calculations

Calculate the EDTA concentration from:

IEDTA] =CTE -

where: W =

KL =

T,, =

B =

weight of standard ZnCl- solution,

molarity of standard ZnCl_ solution used,

volume of EDTA required for ZnCl- determination,i.

volume of EDTA required for blank determination.

F.2.2 Standardization of Bi(N03)_ Against EDTA

Procedure

1. Dilute 100 mL of Bi(N03>3 solutionto 1000 mL with distilled water.

2. Transfer an accurately weighed volumeof 0.01 mol.L"1 Bi(NO3>3 to a 100 mLbeaker containing ^ 50 mL of water.

3. Stir the solution, add 4 drops ofxylenol orange, and adjust the pH to2.5 to 3.0 with the ammonium acetatebuffer and/or dilute HC1.

4. Titrate to a lemon yellow endpoint ,with the now-standardized 0.Q1 mol.L~EDTA solution.

Notes

Aliquot should contain *»» 0.05mmol of Bi.

The pH should be kept in thisrange during the titration.

5. Repeat titration without theBi(N03)3.

To obtain a reagent blank.

Calculations

Calculate the Bi(N0-)_ concentration from:

IBi(NO3)3] -VT

E - VWL

mol/L

- 48 -

where: M = molarity of the standardized EDTA solutionh (see F.2.1)

T_, = volume of EDTA required to titrate sample,El

B_ = volume of EDTA required to titrate blank,

Wn = weight of Bi(N0o). aliquot used.

F.3 Thorium Titration

NOTE: This titration must be carried out in a well-ventilated fumehoodwith adequate shielding against radiation.

F.3.1 Procedure Notes

1. Accurately weigh out from Use borosilicate glassware (Pyrex,Solution A (C.3) a sample containing Kiraax). Soda lime glass may contain2-15 mg Th. calcium which interferes in the

EDTA titration.2. Adjust the pH to < 1.

3. Add 5.000 mL of 0.01 raol.L"1 EDTA. Excess EDTA is added.

4. Add "u 10 drops xylenol orange indicator.

5. Adjust the pH to 2.5 to 2.7 with The pH should be kept in this rangeammonium acetate buffer. during the titration.

6. Titrate the excess EDTA with0.01 mol.L"1 Bi(NO3)3.

F.3.2 Calculations

Calculate the thorium concentration in the fuel solution from:

(Vp \) - (V M ) ,[Thj = E ^ „ 5_B_ 232.04 mg.g X

WF

where: V_ = volume of EDTA added in mL,

M, =* molarity of standardized EDTA solution (see F.2.1),

V- = volume of Bi(N0o)o needed to titrate sample in mL,

KL » molarity of standardized Bi(NO,)3 solution (see F.2.2),

W » weight of fuel solution.r

- 49 -

APPENDIX G

DETERMINATION OF 148Nd BY ISOTOPE DILUTION MASS SPECTROMETRY

G.I Reagents

All reagents should be analytical reagent grade.

1. Di-(2-ethylhexyl)phosphoric acid in 1.5 mol.L~ heptane (HDEHP/heptane).

(a) Dissolve 250 mL HDEHP in an equal volume of n-heptane.

(b) Wash the solution with 500 mL 4 mol.L~ HNO and discard theaqueous phase.

(c) Wash the organic phase with 250 mL 4 mol.L"1 NaOH. If iron ispresent as an impurity, a precipitate of ferric hydroxide will form,suspended1mostly in the aqueous phase.

(d) Discard the aqueous phase and the precipitate. If the organicphase remains cloudy, further washings with HNO and NaOH shouldbe carried out.

(e) Wash the HDEHP twice with 500 mL volumes of saturated ammoniumcarbonate, once with 500 mL 4 mol.L"1 HNO3, and twice with 500 mLof water. The resulting organic phase Is 1.5 mol.L"1 HDEHP inn-heptane.

2. 1.5 mol.L"1 HDEHP/KBrO /10 mol.L"1 HNO .

NOTE 1; Prepare just before use.

NOTE 2; This is used only for the first HDEHP extraction.

Shake 30 mL purified 1.5 mol.L" HDEHP with an equal volume of 10 mol.LHNO- and with 50 mg KBrO,. Discard the aqueous phase.

3. Methanol, anhydrous, A.R. Grade. The methanol must be free from organicimpurities, such as aldehydes.

4. Loading solution (1.56 mol.L nitric acid/80% methanol)Determine accurately the molarity of concentrated HNO3 by titration, anddilute an appropriate volume to 7.80 mol.L"1 with distilled water.Dilute 20.0 mL of the 7.80 mol.L-1 HNO to 100.0 mL with methanol.

5. Eluting solution (0.094 mol.L" nitric acid/80% methanol)Dilute an appropriate volume of concentrated HNO3 to 0.47 mol.L withdistilled water. Dilute 40.0 mL of this solution to 200 mL with methanol.

6. Equilibrated anion exchange resin (BioRad AGMP-1 200 to 400 Mesh, BioRadLaboratories, Mississauga, Ontario).Fill a 3 cm diameter glass column to a height of 10 cm with an aqueousslurry of 200 to 400 mesh AGMP-1 chloride-form resin. Pass 150 mL8 mol.L'1 HNO3 through the resin, followed by 150 mL loading solution.Verify that the resin is free of chloride by adding AgN03 to the finaleffluent, no AgCl should form. Store the equilibrated resin, in loadingsolution, in a glass bottle until needed.

- 50 -

G.2 Apparatus Required

1. Mechanical sample shaker

2. Heat lamp

3. Pumping system capable of maintaining a 0.10 ± 0.01 mL/minute flow ofeluting solution through the ion-exchange columns. A syringe pump,designed to hold up to 4-3Q mL plastic syringes, and delivering 0.10 ±0.005 mL/minute is used in these laboratories.

4. "Chromaflex" disposable plastic ion-exchange columns (KONTES Glass Co.Vineland, J. J., 08360, //EK-420160).

5. Pasteur pipettes.

G.3 Ion-Exchange Column Preparation

Place a small plug of quartz wool into a "Chromaflex" disposable columntip. Insert the tip into a "Chromaflex" tube. Fill the column to aheight of 6.0 cm with equilibrated AGMP-1, 200 to 400 mesh resin. Keepthe resin wet with loading solution. Enlarge the opening of a secondcolumn to •x» 2.5 mm and insert into the top of the column. Push a shortPasteur pipette through the tip for use as a leading funnel (see FigureG.I).

G.4 Separation Procedure

NOTE: This separation procedure must be carried out in a well ventilatedfume hood. Adequate shielding must be provided to limit theradiation exposure to the analyst.

G.4.1 Extraction of Rare Earths

Procedure Notes

1. Take aliquots of fuel solutions A and The sample should contain aboutB (C3, C.4) into clean, new boro- 5-10 mg of heavy elements,silicate screw-capped culture tubes.

2. Add 5oL of KBrO3/HNC>3 conditioned U, Th, Pu and Ce are extractedHDEHP and 3 mg KBrC>3 to each fuel into the HDEHP.solution, and shake vigorously for1 minute.

3. Transfer the aqueous phase to a small Discard the organic phase,disposable plastic beaker and evaporateto dryness under a heat lamp.

- 51 -

Pasteurpipette

Loading/Elutingsolution

AGMP-1resin

Quartz wool

Chromaflex tip

-Chromaflex column

Chromaflex tip

FIGURE G.I: AGMP-1 ANION EXCHANGE COLUMN

T-6-45

- 52 -

Convert the sample to the chloride formby twice evaporating to dryness, andredissolving in concentrated HC1. Re-dissolve in 5 mL of 0.1 mol.L"1 HC1 andtransfer to a clean culture tube.

Add 5 mL of 1.5 mol.L~ HDEHP (notconditioned with KBrOj) to the sampleand shake for two minutes. Transferthe organic phase to a clean culturetube, and repeat the wash of theaqueous phase with another 5 mL ofHDEHP.

Wash the HDEHP twice with 5 mL volumesof 1 mol.L"1 HNO3, and transfer theaqueous phase to a plastic beaker.Evaporate the samples to dryness undera heat lamp.

Separate the Nd from other rare earthsby ion-exchange, as described inSection G.4.2, below.

Nd and other rare earths areextracted into the HDEHP. Cs andBr residue remain in the aqueousphase.

Nd and other rare earths areextracted into the HNO».

G.4.2

1.

2.

3.

Ion-Exchange Separation of Nd from Rare Earths, Am and Cm

Procedure Notes

Prepare an AGMP-1 column as describedin Section G.3. Place a waste recep-tacle under the column.

Dissolve the sample in 0.5 mL loadingsolution and transfer to the column.

Rinse the sample container with 0.5 mLloading solution, and transfer to thecolumn.

6.

7.

Pass another 2.0 mL of loading solu-tion through the column.

As soon as the column reservoirempties, add 1 mL of eluting solutionto the column, remove the Pasteurpipette, and immediately connect theline from the eluting solution pumpto the column, as shown in Figure 7.

Pump 12 mL of eluting solution throughthe column. Discard this fraction.

Place a clean plastic beaker, labelled"Nd fraction" under the column, andcollect the next 10 mL of eluent.

Fission products other than rareearths are eluted.

The volume of eluting solutionabove the resin should remain ata constant level when the line isconnected and an air-tight sealis formed.

Rare earth fission products heavierthan Nd are eluted.

Nd is eluted.

- 53 -

8. Discard the ion-exchange column. The column is radioactive. Dispose inan appropriate way.

9. Evaporate the neodymium fractionto dryness.

10. Dissolve the neodymium fraction in as small a volume of 1 mol.L~ HNO,as possible and evaporate the solution onto a tantalum mass spectro-meter filament by passing a low voltage current through the filament.

11. Mount the filament into the mass spectrometer and determine the Nd/l^Nd ratio using the technique described by Shields 0-4) #

G.5 Calculation of Nd Concentration in the Fuel Solution

The Nd concentration in the fuel solution is calculated using thefollowing equation:

where: INR] * atoms of Nd per gram of fuel solution,142 142

IN.J = atoms of Nd per gram of Nd tracer solution

(obtained from B.I.7).,

W = weight of tracer solution in grams,

W " weight of fuel solution in grams,Nd/ Nd isotopic ratio in purified neodymium fraction

of the (fuel solution/l^Nd tracer solution)mixture,

4 = 148Nd/142Nd isotopic ratio in the 142Nd tracer solution(obtained fron> B.I.6),

4 - 148Nd/142Nd isotopic ratio in the purified neodymiumfraction of the unspiked fuel solution.

Example:

W T N = 0.604 g

W_ = 1.Q69 g16 142

JN»] = 1.38x10 atoms Nd per gram of tracer solution

4 " 0.4623

4 * 0.00058

4 - 63.35,„ , (0.604)(1.38xl016) [0.4623-0.0005811 N8 J " 1.069 " [1-0.4623/63.35J

(7.80xl0:L5)CQ.4651) - 3.63xlO15 atom 148Nd per gram fuel solution

- 54 -

APPENDIX H

BURNUP CALCULATION

The universal burnup calculation using Nd as burnup indicator is:

INJ n aBU(GJ/kg HE) = ° "

*|/Q

where: IN] = Nd in atoms per gram fuel solution

£HEj = concentration of heavy element per gram fuel solution

i = fraction of total fissions occurring in ith fissioning isotopek. = atoms Nd produced per Joule of energy produced in the

fission process

Example: INg] = 3.63x10 atoms Nd per gram fuel solution

IHE] = ]U] + fThJ = 36.83 yg U + 1Q.0 mg Th/g fuel solution

«233= °'2

0(235 = °'8 8 148k233 = 4- 0 5 x l° atom Nd Per Joule (see Table 3)

k235 = 5 * 2 5 x l ° 8 atom 148Nd Per Joule

BU(GJ/kg HE) = 3' 6 3 x l°(0.0368 + 10)( 0.2 + 0.8 \

4.05xl08 5.25xlO8 J

7.29xl05 J/mg HE = 729 GJ/kg HE

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