Fission Product Yields of 233U, 235U, 238U and 239Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons

Nuclear Data Sheets 111 (2010) 2965–2980 www.elsevier.com/locate/nds

Fission Product Yields of 233 U, 235 U, 238 U and 239 Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons J. Laurec∗ , A. Adam† , T. de Bruyne‡ , E. Bauge§ , T. Granier, J. Aupiais, O. Bersillon† , G. Le Petit Commissariat ` a l’Energie Atomique, Centre DAM-Ile de France (CEA DAM DIF), 91297 Arpajon, France

N. Authier, P. Casoli Commissariat ` a l’Energie Atomique, Centre de Valduc, 21120 Is-sur-Tille, France (Received 5 July 2010; revised received 7 September 2010; accepted 1 October 2010) The yields of more than fifteen fission products have been carefully measured using radiochemical techniques, for 235 U(n,f), 239 Pu(n,f) in a thermal spectrum, for 233 U(n,f), 235 U(n,f), and 239 Pu(n,f) reactions in a fission neutron spectrum, and for 233 U(n,f), 235 U(n,f), 238 U(n,f), and 239 Pu(n,f) for 14.7 MeV monoenergetic neutrons. Irradiations were performed at the EL3 reactor, at the Caliban and Prospero critical assemblies, and at the Lancelot electrostatic accelerator in CEAValduc. Fissions were counted in thin deposits using fission ionization chambers. The number of fission products of each species were measured by gamma spectrometry of co-located thick deposits. Contents

I. Introduction II. Definitions, principles A. General experimental method B. Fission yield: definitions 1. Constant flux 2. Sequential irradiation 3. Case of a filiation III. Experimental procedure A. Principle B. Characteristics of the irradiation facilities 1. Thermal neutrons 2. Fission neutrons 3. 14 MeV neutrons C. Samples and mass measurements 1. Fissile materials 2. Sample fabrication 3. Mass measurements D. Measurement of the number of fissions 1. Fission chamber efficiency 2. Electronics and data acquisition E. Fission product measurements

2965

1. 2. 3. 4.

Gamma detectors Electronics Detector efficiency Gamma spectra analysis

2966 2966 IV. Fission product yields A. Uncertainties and data updates 2966 1. Measurement combination 2966 2. Uncertainty in fission yields 2966 B. Thermal spectrum 2967 C. Fission spectrum 1. 233 U 2967 235 2. U 2967 239 3. Pu 2967 D. 14 MeV neutrons 2967 1. 233 U 2968 2. 235 U 2968 3. 238 U 2968 4. 239 Pu 2968 2969 V. Conclusion 2969 2969 Acknowledgments 2969 2970 References 2970 I.

∗ 1940-1998

In memoriam retired ‡ Now in CEA DEN Saclay § Corresponding author, electronic address: [email protected] † Now

0090-3752/$ – see front matter © 2010 Published by Elsevier Inc. doi:10.1016/j.nds.2010.11.004

2970 2970 2970 2971 2971 2971 2971 2971 2972 2974 2974 2975 2977 2978 2978 2978 2978 2978 2978 2979 2980

INTRODUCTION

Reference data on fission product yields for major actinides are essential for nuclear research as well as for many industrial applications such as assessments of fis-

Fission Products Yields of

233

U,

235

U, . . .

sion energy, combustion rate, neutron flux or safety applications. In particular, fission product yields are used to determine the number of fissions occurring in samples of uranium and plutonium irradiated by neutrons. While a large amount of experimental data is available for thermal neutrons, they are much scarcer for fission neutrons and 14-MeV neutrons. Up to now, the most precise techniques for measuring fission product yields are radiochemistry and mass spectrometry, even though they do not apply equally well for all fission products. In this context, an experimental program using the Prospero and Caliban [1] critical assemblies and the Lancelot 14-MeV neutron generator was performed in the mid-seventies onward by CEA DAM teams. The experiments were based on the fission chamber method to determine the number of fissions in a thin deposit and on gamma spectrometry to measure the number of nuclei of each species produced in a thicker target. The masses of the irradiated samples were measured by mass spectrometry and alpha spectrometry. This approach is essentially the same as the one used in the ILRR program [2]. These experiments have produced very valuable results on fission product yields for 233 U, 235 U, 238 U, and 239 Pu in neutron fields of three ranges of incident neutron energy: thermal spectrum, fission neutron spectrum from a fast critical assemblies and 14.7 MeV. These results were initially published in CEA DAM internal reports [3–5] and in a CEA report [6], all in French. These very well documented reports are not widely available today and it was found that a fresh publication gathering the results and relevant information would be useful. This also gives the opportunity to update the results of the fission product yields using the latest nuclear data (intensities and half-lives). Clarification is also given on the associated uncertainties.

be precise, this measurement requires a gamma activity corresponding to about 1011 fissions which far exceeds the number of fissions obtainable in a 10 μg/cm2 -thick sample. For that reason, a second sample of the same fissile material but with a much larger mass and thickness (greater than 1 mg/cm2 , with a total mass ranging from 1 to 6 mg), is associated with the fission chamber. This target has the same geometry as that of the deposit inside the chamber. It is stuck to it and irradiated simultaneously. The neutron flux experienced by the two samples are thus as close as possible. Nevertheless a small adjustment of the neutron flux is usually required. It is determined by the use of thin activation foils placed at judicious locations. The use of two deposits of different masses implies an additional difficulty: the measurement of the number of fissions in the sample is not direct anymore. To obtain it, it is then necessary to know the mass ratio of the two samples very precisely. The number of fissions Nf induced in the sample by time unit is then given by the following relationship N f = nf

M r, m

(2)

where nf is the fission count rate from the fission chamber, m is the mass of the thin deposit inside the chamber, M is the mass of the thick sample outside the chamber and r is the coefficient for neutron flux adjustment. B.

Fission yield: definitions 1.

Constant flux

During an irradiation time interval dθ the variation of the number of nuclides Ni can be written as

II.

DEFINITIONS, PRINCIPLES

dNi = Yi Nf − λi Ni , dθ

A.

General experimental method

where Nf is the number of fissions per time unit and λi is the decay constant for nuclide i. Integrating this equation for an irradiation duration θ gives

The fission yield of a product nuclide i is given by the ratio of the number of atoms Ni of nuclide i formed in the sample by the number of fissions Nf having occurred in that sample during irradiation Yi =

J. Laurec et al.

NUCLEAR DATA SHEETS

Ni . Nf

(1)

In order to measure the number of fissions induced in the sample, the fission chamber method has been used. In this method, the sample is a thin actinide deposit (on the order of 10 μg/cm2 ) which covers one electrode of the ionization chamber. Low thickness is required so that the fragments can escape the sample and ionize the pressurized gas, inducing an electric signal. Gamma spectrometry is used to measure the number of produced nuclides of each species of interest. However, to

Ni = Yi Nf

1 − e−λi θ , λi

(3)

(4)

where Ni is the number of atoms remaining at the end of irradiation. In the case of the fission chamber method one gets Yi = Ni

λi m 1 1 . 1 − e−λi θ M nf r 2.

(5)

Sequential irradiation

In the case where the irradiation corresponds to a successions of n irradiation periods with constant flux and

2966

Fission Products Yields of

233

U,

235

U, . . .

characterized by a duration θj and numbers of fissions Nf j , then the number of atoms present at measurement time Nimeas is given by nmeas = Yi i

n 

Nf j ·

j=1

−λi θi

1−e λi

e−λi

n   k=j+1

J. Laurec et al.

NUCLEAR DATA SHEETS

θk +

n 

with n 

Θ=

 tk ,

θk +

k=j+1

meas NCi = YCi

 +

NCmeas = NC e−λC t + NB

λB (e−λB t − e−λC t ), λC − λB

(7)

where t is time elapsed since the end of irradiation. The variation dNC of the number of atoms C during a time interval dθ of irradiation is dNc = λB NB − λC NC . dθ

(8)

Since YB Nf (1 − e−λB θ ) , λB

(9)

it follows dNC = YB · Nf (1 − e−λB θ ) − λC NC , dθ

(10)

and  Nf  λC (1 − eλC θ + (e−λB θ − e−λC θ ) . λC λB − λC (11)

n  j=1

NBij

1 − e−λCi θj e−λBi θj − e−λCi θj + λCi λBi − λCi

III.



 e−λCi Θ .

(14)

  λBi e−λBi Θ − e−λCi Θ λCi − λBi

+ NCij e−λCi Θ ,

(12)

EXPERIMENTAL PROCEDURE A.

Principle

Samples of known mass are irradiated and gamma spectrometry is performed off-line to count the number of nuclei of each species of interest present in the sample. Nuclear data is then used to determine the number of formed nuclei at the end of irradiation. The number of fissions having occurred in the sample during the irradiation is measured using the fission chamber technique. The method employs two sub-samples made from the same actinide sample. One is a thin deposit which is placed within a fission chamber. The latter delivers an electric signal when a fission occurs within the deposit. The second, thicker sample is placed very close to the fission chamber so that the neutron flux experienced by both samples is as close as possible. In practice, small flux corrections determined using activation measurements of reference foils may apply. The ratio of the masses of the two samples need to be known to a high precision. This is assured by coulometry, mass spectrometry and alpha counting measurements of the samples. A typical fission chamber used in the experiments is shown in Fig. 1. A target setup comprising a thick sample, a fission chamber and activation foils is shown in Fig. 2. For irradiations performed outside the critical assemblies (see Section III B 2), this setup was surrounded by a 0.8 mm thick Cd shielding which proved to reduce the thermal neutron flux by a factor of 5×10−4.

B.

In the general case of a sequential irradiation with variable flux, each i chain yields meas NCi =

 e−λBi Θ − e−λCi Θ Nf j (1 − e−λBi θj ) + λCi − λBi j=1

n 

Case of a filiation

If the half-life of the precursor B is not negligible with respect to that of the measured nuclide C, it must be taken into account. It is considered in this case that the independent yield of daughter C is negligible compared with the cumulated yield at the precursor level. This approximation is always justified for the fission products studied in this paper. If NB and NC are the number of atoms of the nuclides B and C at the end of an irradiation time θ, then at tmeas the time at the beginning of the measurement one gets

NC = YB ·

(13)

It then follows that

k=j

with Nf j being the number of fissions per time unit during sequence j. If the flux varies during the sequences of irradiation, then this variation can be discretized to a good approximation if the time step is small compared to sequence duration, and equation 6 can be applied.

NB =

tk .

k=j

(6)

3.

n 

Characteristics of the irradiation facilities 1.

Thermal neutrons

Irradiations have been conducted at the thermal column of the EL-3 heavy water-cooled reactor (decommissioned in 1982) at the CEA in Saclay [7]. The neutron flux at the irradiation point was 109 neutron/cm2 /s.

2967

Fission Products Yields of

233

U,

235

U, . . .

J. Laurec et al.

NUCLEAR DATA SHEETS

Prop. to n Fluence/MeV

1.25 MeV whereas outside the assembly it is about 0.75 MeV [8]. Both critical assemblies are installed in a large cell. Characteristic neutron spectra simulated in both assemblies using the Tripoli-4 [9] code are given in Fig. 3. The neutron flux was 1011 neutron/cm2 /s within the cavity and 1010 n/cm2 /s outside the assembly.

FIG. 1: Schematic view of the radial fission chamber used for the determination of the number of fissions.

10 10 10 10 10

-3 -4 -5 -6

Prospero-U

-7

Prop. to n Fluence/MeV

10

10

-1

1 10 Neutron Energy (MeV)

10 10

-1

-2

-3

Caliban -4

10

2.

10

1

10

FIG. 2: Target setup used in the irradiations consisting of the fission chamber (1), the thick actinide activation target (4), the thin actinide deposit (5). activation foils (3) and zircaloy rings (2).

-2

-2

10

-1

1 10 Neutron Energy (MeV)

FIG. 3: Neutron spectral shapes outside the Prospero-U assembly (top panel) and in the cavity of the Caliban assembly (bottom panel) simulated using the Tripoli-4 code.

The Prospero-U and Caliban assemblies are still up and running today.

Fission neutrons

The irradiations have been performed at the Prospero and Caliban [1] critical assemblies at the CEA facility in Valduc. Caliban has been in operation since 1970. It is a compact experimental pulsed-type reactor with a highly enriched metallic uranium core. The objects to be irradiated are inserted either in the central cavity or outside the assembly. The mean neutron energy in the cavity is 1.35 MeV. Outside the assembly, near the core, the mean neutron energy remains essentially the same [8]. The Prospero reactor has been in operation since 1968. It operates at continuous power. Prospero was modified in 1975 at the time of the experimental campaigns reported here. The original plutonium core was replaced by a core made of highly enriched metallic uranium. This change has resulted in slightly softer spectrum. In the two versions of the Prospero assembly (denoted as ’Prospero-Pu’ and ’Prospero-U’ in this article) the core is surrounded by a reflector made of depleted uranium and stainless steel. The objects to be irradiated can be placed either in the central cavity or outside the assembly. The mean neutron energy within the cavity is about

3.

14 MeV neutrons

Irradiations at 14.7 MeV have been performed using the now dismantled Lancelot electrostatic accelerator at CEA-Valduc. Neutrons were produced using the T(d,n) reaction by bombarding a tritiated target with accelerated deuterons at a 160 KV voltage. The target was a rotating TiT water-cooled target. Neutron flux at 2 cm downstream from the production target, on the beam axis, was on the order of 5×1010 neutron/cm2 /s.

C.

Samples and mass measurements 1.

Fissile materials

For each element, isotopic analyses were performed (see Table I). Radiochemical separations were carried out for 239 Pu in order to remove 241 Am. The isotopic purity of 235 U was 99.9% and this isotope was used without further

2968

Fission Products Yields of

233

U,

235

U, . . .

purification. The isotopic analysis of 233 U revealed the presence of small amounts of 232 U which leads to numerous additional γ-rays due to the presence of 232 Th and its daughter nuclides. A radiochemical separation Th/U was performed before the target fabrication. TABLE I: Isotopic composition of fissile actinide samples. Sample 239 Pu 233 U 235 U 238 U

40/39 1.71×10−2 34/33 4.2×10−3 34/35 7.0×10−6 34/38 9.0×10−6

2.

J. Laurec et al.

NUCLEAR DATA SHEETS

Isotopic ratio 41/39 3.7×10−4 38/33 1.8×10−2 32/33 2.3×10−6 38/35 9.0×10−6 35/38 2.0×10−6 36/38 5.4×10−6

Sample fabrication

As mentioned above, two types of actinide samples were made: thin deposits for the fission chambers and thicker ones for fission product analyses. The thinner samples, mounted in the fission chambers, were made by electroplating. This technique ensures strong adhesion onto titanium or zircaloy backings. The electroplating consists in depositing the actinides from an aqueous solution of 5.5 M NH4 Cl at pH = 1 (Pu) or 2 (U) at a constant density of charge inside the cell (e.g. i = 1.5 A, V < 20 V). The duration does not exceed 30 minutes. After the electroplating, the sources were carefully washed with alcohol and water. Finally, the actinide deposit is heated in order to fix it onto the Ti or zircaloy backing. The final form is an oxide: U3 O8 or PuO2 . For the thick samples, another technique was necessary because electroplating is not convenient for depositing aerial masses exceeding 100 μg/cm2 . Electrospraying was chosen because it can provide a very good homogeneity of the deposit [10, 11]. In addition, this technique is convenient for making a series of samples of the same thickness. Consequently, very precise relative measurements can be performed between targets of the same series. A solution of the actinide, generally a salt, dissolved in a convenient solvent for electrospray (methanol, ethanol, acetone etc.) passes through a glass capillary (internal diameter ∼400 μm) containing an electrode (diameter 390 μm) positively charged at a nominal voltage depending on the nature of the solvent (i.e. 4000 - 10000 V, intensity 1 mA). The micro-droplets explode into a very thin aerosol due to Coulomb forces in such a way that the solvent is evaporated before reaching the substrate. The charged particles are accelerated before impacting the substrate. The distance between the capillary nozzle and the support is about 1 cm. The substrate is an aluminum disk of 0.1 mm thick. Aluminum is chosen because it can be easily chemically dissolved to perform mass measurement and radiochemical separations while its neutron activation produces non-interfering gamma emitter, short-lived products. Because of a 6 MeV threshold, only a very weak activation by fission neutrons is expected.

3.

Mass measurements

For the thin targets, the mass was systematically measured by α spectrometry in a well defined geometry. Comparisons with mass spectrometry were performed for a few targets, before irradiation. Results were in excellent agreement as shown in Table II. Mass determination by thermo-ionisation mass spectrometry (TIMS) was performed after dissolving the target (actinide and support) in an acidic medium and using the isotopic dilution method (242 Pu) for quantitative determination. After several irradiations, a few chambers were dismounted and the samples recovered. New mass spectrometry measurements and α-spectrometry confirmed the first determination by α-spectrometry. For the thick targets, the mass determination by α-spectrometry is not possible and the IDMS (isotopic dilution mass spectrometry) is the only precise method. However, when possible, comparisons with γ-spectrometry (for γ-emitting actinide isotopes) or coulometry (235 U) were achieved after dissolving a few targets. All results were in excellent agreement, i.e. within the uncertainty range (0.5 - 1.0 %). TABLE II: Comparisons between mass determinations of 235 U, 238 U and 239 Pu thin samples by α spectrometry in defined geometry and thermo-ionisation mass spectrometry (TIMS). Sample Chamber # α- spect. (μg) Mass spect. (μg) 239 Pu 1577 2.23±0.06 2.26± 0.02 233 U 1837 37.4±1.0 37.6±0.5 235 U 1418 5.36± 0.10 5.49±0.03

D.

Measurement of the number of fissions

The fission chambers were fabricated at the CEA in Saclay and Cadarache. Their metal parts were made of titanium and zircaloy in order to minimize neutron absorption. Insulating parts were made of ceramics: steatite, and sintered alumina. The fission chambers were filled with argon at a 12 bar pressure. Depending on the chamber, external diameter was 12 or 20 mm. Fissile material was deposited onto thin foils of titanium or zircaloy. Thickness of the deposits was of a few microgram/cm2 for a diameter between 8 and 14 mm depending on the chamber type. Nominal voltage bias was between +200 and +300 Volts. Fission chambers were built according to two designs. One with a coaxial output and the other with a radial output. The first type, which was used for most measurements, is shown in Fig. 1.

1.

Fission chamber efficiency

The thickness of the fissile deposits in the chambers was always less than 10 μg/cm2 . Therefore, the absorp-

2969

Fission Products Yields of

233

U,

235

U, . . .

tion of the fission fragments is negligible considering the range of fission fragments in uranium oxide or plutonium oxide [36]. A 100% efficiency can thus be assumed. This assumption was experimentally confirmed in two ways. Firstly, the alpha activity of the sample in the chamber was measured by counting the alpha-induced signals from the fission chamber. This activity was found to be in excellent agreement with that measured before mounting the sample in the chamber by alpha spectrometry. Secondly, three fission chambers were assembled including very thin gold and cobalt foils of known mass. These foils were inserted between the chamber case and the backing of the fissile deposit or placed outside of the chamber, against the chamber case. The chambers were then irradiated in a thermal neutron flux. The neutron integrated flux experienced by the gold and cobalt foils was determined by measuring their activities, taking into account the associated capture cross sections. This measurement was then compared to the flux obtained by counting the number of fissions signals occurring during the irradiations. Once again, the agreement was extremely good confirming the validity of the total efficiency of the chamber. Also, relative measurements of the number of fissions were performed for several fission chambers irradiated over the same integrated flux of thermal neutrons. The number of fissions per mass unit of fissile material per second obtained were in very good agreement.

2.

Electronics and data acquisition

The signals out of the fission chamber are sent to a multichannel analyzer and to two electronic scalers running alternatively which were recorded onto perforated strips using a teletype. In this way the integral of the number of fissions and the flux variation as a function of time during irradiation was obtained. A discriminator was used to reject signals due to electronic noise and alpha particles.

E.

J. Laurec et al.

NUCLEAR DATA SHEETS

Fission product measurements

The determination of the number of atoms for each fission product was performed by γ-spectrometry of the irradiated targets. For the fission products of low activity (144 Ce, 155 Eu, 156 Eu), radiochemical separations were necessary to improve the precision. The separation of individual rare earths is based on isotopic exchange with a stable carrier, coprecipitation of the corresponding hydroxide, group separation by extraction chromatography (HDEHP) and ion exchange (Dowex 2×8), and finally selective elution from a cationic ion exchange (Dowex 50×12) by a concentration gradient of ammonium lactate. The chemical yield is determined by gravimetry after precipitation by oxalic acid and heating the precipitate until the formation of the stoichiometric oxide CeO2 or Eu2 O3 .

The γ spectrometry chains were comprised of a GeLi detector, an electronic module for signal amplification and shaping, an analog-to-digital converter (ADC) and a data acquisition system.

1.

Gamma detectors

The Ge-Li detectors used were of planar or coaxial geometry with volumes on the order of 1 cm3 (planar) or 40 cm3 (coaxial). The energy resolutions ranged from 0.6 to 0.9 keV at 122 keV and from 1.9 to 1.33 keV at 1.33 MeV, depending on the detector. The input stage of the preamplifiers was mounted in the cryostat and cooled down in order to reduce electronic noise. The detectors were shielded with 10 cm lead in order to reduce the level of gamma-ray background.

2.

Electronics

The amplifiers were either Ortec 472 or Enertec 7129. The shaping times were chosen to obtain the best energy resolution since the counting rate was always low. They varied from 2 to 4 μs depending on the detector. The ADCs were Intertechnique CT103. They were used with a Plurimat N20 acquisition system.

3.

Detector efficiency

Great care was taken in order to determine the efficiency curves of the detectors. They were crosschecked in the framework of inter-comparison programs with other laboratories. The standard sources used came from LMRI (CEA-Saclay) or Amersham. These standards (51 Cr, 54 Mn, 57 Co, 60 Co, 85 Sr, 88 Y, 109 Cd, 113 Sn, 137 Cs, 139 Ce, 141 Ce, 144 Ce, 203 Hg, and 241 Am cover the energy range from 60 keV to 1836 keV. In addition, a multigamma standard (152 Eu) was used. These primary standards enabled determination of a precise efficiency curve in the quasi-linear region above 150 keV. In order to improve the curve below 150 keV, multigamma sources with gamma lines from 60 to 400 keV were used (166 Hom , 169 Yb, 182 Ta) with the relative intensities given in Ref. [12]. The distances between the sample and the detector were 10, 20 or 50 cm. They were chosen to obtain low count rates (< 100 counts per second) and to avoid the creation of sum peaks in the spectra. The efficiency curve was obtained by smoothing the experimental points using a polynomial. Solid angle correction was applied to account for the differences between point-like and extended sources. The overall uncertainty on the efficiency taking into account solid angle correction was estimated to 2.0% (2σ).

2970

4.

233

235

U, . . .

NUCLEAR DATA SHEETS

Gamma spectra analysis

IV.

FISSION PRODUCT YIELDS

A.

Uncertainties and data updates

Fission Products Yields of

U,

Most of the fission products were measured by performing gamma spectrometry directly on the target. Nuclides measured in this way are: 95 Zr, 97 Zr, 99 Mo, 103 Ru, 105 Rh, 127 Sb, 131 I, 132 Te, 133 I, 135 Xe, 136 Cs, and 140 Ba. 144 Ce has been measured after chemical isolation of cerium to compensate for its low activity due to a relatively low gamma branching ratio (∼11%) and a rather long half-life (∼285 days). In this measurement however, the chemical yield value is not needed since an adjustment can be done on 141 Ce which is measured on the same sample. For all fission products, the most intense gamma lines are used. Most of them are well isolated in the gamma spectra. The first measurement is performed 24 hours after the end of irradiation in order to let the short halflife nuclides decay. The spectra analyses were realized using a dedicated computer program. The background below the photopeaks was assumed linear, its level being determined using several channels on both sides of the peak. The program makes it possible to resolve well defined doublets. The same program was also used for the efficiency determination. The fissile nuclides 233 U, 235 U and 239 Pu are gamma emitters and in some cases their gamma lines interfere with those of the fission products. For instance, in the case of 239 Pu, the lines at 144 keV and 146 keV interfere with that at 145 keV from 141 Ce and the most intense 239 Pu line at 129 keV (0.0063 %) interferes with the measurement of the 133 keV line from 144 Ce. In order to overcome this problem, the targets were measured before irradiation in the same geometrical conditions as for the fission products. In this way, the contribution from the fissile materials can be subtracted from the fission product spectra. In the case of the 233 U targets, this correction is complicated by the presence of 232 U. The relative statistical uncertainty associated with the counting of a gamma peak signal is computed as

√ ΔS S T + SB =2 , S S

(15)

where S is the number of counts in the photopeak exclusive of background, SB is the number of background counts situated below the peak and ST = S + SB is the total number of counts in the peak energy region. Use of this formula implies that statistical errors are given with a 2σ confidence level. In addition to the statistical uncertainty, a deconvolution uncertainty depending on the spectrum configuration in the region of the considered photopeak was taken into account. The value of this uncertainty is 2% at maximum, again with 2σ confidence level.

J. Laurec et al.

1.

Measurement combination

The results on the fission product yields for a given actinide in a given neutron field were obtained from several experiments as described in the next section. Moreover, for each experiment several gamma spectrometry measurements were performed leading, in general, to several measurements for each fission product. The final value for each fission product is the combination of these different measurements through weighted averages, the weight being defined as the inverse of the square of the statistical uncertainty associated with each individual measurement.

2.

Uncertainty in fission yields

The different terms contributing to the total uncertainty associated with the fission product yields are listed in Table III. The systematic terms have been estimated as constants for all the experiments. We have chosen to keep the procedure followed in the original report (Ref. [6]). Additional uncertainties come from the nuclear data which are used to determine the number of products from gamma counting: gamma intensities and nuclide half-life. The values of these nuclear constants have sometimes varied significantly over time since the experiments. For a proper use of the fission product yield data of this work, an update needed to be performed using the best nuclear data knowledge as of today. The nuclear data constants which we used to obtain the results are taken from the NUDAT 2.5 data base as of April 2010. Table IV gives the gamma lines which were used in determining the fission product yield and the corresponding nuclear data. In the cases of a single gamma line, the yield update is obtained by multiplying the yield quoted in Ref. [6] by the ratio of the gamma intensity originally used to that from the latest evaluation. The same applies for decay constants. For some fission products two gamma lines have been used in the gamma spectrometry measurements. In these cases the statistical weight of each gamma line in the measurement is required to update the fission yield. Since the information was not present in the original references, we estimated these statistical weights on the base of an original 235 U counting sheet from these experiments. That estimation was confirmed by a coarse analysis of the gamma spectrum measured with Ge-Li detector on 238 U irradiated in a 14.7 MeV neutron flux (shown in Fig. 4). The update is then given by  I old I2old  1 Y new = Y old w1 new , + w2 new I1 I2

2971

(16)

Fission Products Yields of

233

U,

235

U, . . .

NUCLEAR DATA SHEETS

where w1 and w2 are the normalized weights corresponding to each gamma line. These weights are given in Table V.

TABLE V: Statistical weights used to update the yields in the case of the yield measurements being based on two gamma lines (see text). Nuclide γ line Weight, % 95 Zr 724 keV 50 757 keV 50 97 Zr 658 keV 50 743 keV 50 127 Sb 473 keV 50 686 keV 50 136 Cs 818 keV 60 1048 keV 40 140 Ba 163 keV 50 537 keV 50 147 Nd 91 keV 50 531 keV 50 155 Eu 86 keV 50 105 keV 50 156 Eu 89 keV 90 646 keV 10

TABLE III: Terms contributing to the uncertainty associated with the yield values. The values are given for a 2σ confidence level. The uncertainties associated with nuclear data (i.e., γ intensities and half-lives) are not included. Source of uncertainty Stat. unc. Syst. unc. Photopeak area 0.3 - 6 % γ detection efficiency 2.0 % Samples’ mass ratio 1-2% 0.8 % Meas. of number of fissions 0.5 % Neutron flux readjustment 0 - 1.5 % -

TABLE IV: Nuclear data used to obtain the fission product yields: gamma-ray energies and intensities, half-lives and associated uncertainties (1σ confidence level). References are given in the last column. Nuclide 95 97

Zr

Zr (97(m) Nb)

99

Mo (99m Tc) 103 Ru 105 Rh 106 Ru (106 Rh) 127 Sb 131

I I 132 Te 135 Xe 136 Cs 133

137

Cs (137m Ba) 140 Ba 141

Ce Ce 144 Ce 147 Nd 143

155 156

Eu Eu

Eγ keV 724.2 756.7 657.9 743.4 140.5 497.1 318.9 621.9 473.0 685.7 364.5 529.5 228.2 249.8 818.5 1048.1 661.6 162.7 537.2 145.4 293.3 133.5 91.0 531.0 86.5 105.3 89.0 646.3

B.

Iγ % 44.27 54.38 98.23 93.09 89.06 91.0 19.1 9.93 25.8 36.8 81.5 87.0 88 90 99.70 80 85.10 6.22 24.39 48.29 42.8 11.09 28.1 13.37 30.7 21.10 8.4 6.3

u(Iγ ) T1/2 days 0.22 64.032 0.22 0.08 0.698 0.16 0.24 2.747 1.2 39.26 0.6 1.473 0.23 371.8 1.6 3.850 2.0 0.8 8.025 3.0 0.867 3 3.204 3 0.381 13.04 3 0.2 10979.2 0.09 12.753 0.22 0.20 32.508 0.4 1.376 0.19 284.91 0.5 10.98 1.1 0.3 1734.84 0.5 1.1 15.19 0.5

u(T1/2 ) Ref. days 0.006 [16] 3×10−4 [17] 4×10−4 0.02 0.002 1.8 0.05

[18] [19] [20] [21] [22]

0.001 0.004 0.013 0.001 0.03

[24] [23] [25] [26] [27]

32.85 0.002

[28] [29]

0.013 0.002 0.05 0.01

[30] [31] [32] [33]

5.11

[34]

0.08

[35]

Thermal spectrum

The goal of these experiments was to validate the experimental procedure by taking advantage of the relatively good knowledge of the fission product yields in the

J. Laurec et al.

case of 235 U in a thermal neutron field. In Ref. [6] the results on the fission product yields are reported to be in good agreement with the reference data at that time. We have repeated this comparison with modern data, namely from the JEFF-3.1 [13] and ENDF/B-VII.0 [14] evaluations. The experimental results for 235 U are given in Table VI together with the evaluated data. The comparison reveals that our results are in very good agreement with the evaluations taking account of the uncertainties (with exceptions discussed below). For 97 Zr, the relative differences (defined as 1.0−M/E, with M our measured yield, and E the evaluated one) are larger but close to our measurement uncertainty. For 103 Ru and 140 Ba, our measurements disagree (outside 1σ confidence level) with the value of JEFF-3.1 but are found to be in good agreement with those of ENDF/B-VII.0. Conversely, our measurement of the 105 Rh 235 U thermal fission yield is inconsistent with the ENDF/B-VII.0 value, but consistent with that of JEFF-3.1. In the case of 147 Nd, the agreement is not as good with both evaluated values. Since there was no special difficulty associated with the gamma counting of that particular fission yield, one might speculate that the nuclear data (gamma intensities and half-lives) used to normalize the number of atoms could be responsible for this relative disagreement. Finally, looking at the relative differences shown on Table VI reveals a trend where our measured yields seem to almost systematically underestimate the evaluated yields by a little more than 1%, while staying within uncertainties. Since the very same procedure was used for measuring yields in fast critical assembly spectrum and for 14.7 MeV neutrons, we expect these other measurements to reach a level of quality comparable to that exhibited by the above

2972

Fission Products Yields of

233

U,

235

U, . . .

NUCLEAR DATA SHEETS

J. Laurec et al.

FIG. 4: Typical gamma counting spectrum acquired with a 238 U target irradiated in the 14.7 MeV neutron flux of Lancelot. The Ge-Li detector intrinsic efficiency corresponds to a 10% efficiency ratio at 1.333 MeV relative to a 3 in. × 3 in. NaI(Tl) cylindrical crystal (manufacturer’s specification).

comparison between our measured 235 U thermal neutron fission yields and current evaluated values. Another thermal spectrum measurement has been performed for 239 Pu (reported as a test in Ref. [3] but omitted in Ref. [6]) with experimental conditions very similar to those of the thermal 235 U measurements. The results of these measurements are reported on Table VII, and compared with current evaluated data from the JEFF3.1 [13] and ENDF/B-VII.0 [14] libraries. Clearly, 239 Pu thermal neutron fission yields are less well known than those of 235 U, as evidenced by the larger spread between evaluated values from the JEFF-3.1 and ENDF/B-VII.0 libraries. Nevertheless, such comparison yields insights into the systematic uncertainties associated with our measurements. Fig. 5, which shows the ratios between our measured thermal neutron fission yields for 235 U (top panel) and 239 Pu (bottom panel), and the evaluated values from the JEFF-3.1 library and the ENDF/B-VII.0 library, reveals similar behaviour for both 235 U and 239 Pu ratios as

a function of fragment mass. These remarkable similarities are the manifestation of systematic errors that can be attributed to a hard to disentangle combination of uncertainties associated with energy-dependent gamma counting efficiencies and uncertainties associated with the nuclear data (half lives and gamma intensities which are isotope-dependent). On the other hand, while our 235 U data seem to exhibit a trend towards underestimation of evaluated values by 1%, our 239 Pu data exhibit a trend towards overestimation of evaluated yields by as much as 2.5%. It is diffcult to identify the cause of these opposing trends, however they provide us with empirical evidence that the isotope-independent relative uncertainties associated with our measurements can be of the order of 1 to 2.5%. Finally, the agreement between our thermal neutron fission yields and the associated evaluated values can be quantified by calculating the χ2 between their ratios and unity. For the 235 U target, the average χ2 per degree of freedom (number of measured points) amounts to 0.84 for

2973

Fission Products Yields of

233

U,

235

U, . . .

J. Laurec et al.

NUCLEAR DATA SHEETS

TABLE VI: Results of fission product yield measurements for 235 U in a thermal neutron flux compared with JEFF-3.1 [13] and ENDF/B-VII.0 [14]. Relative uncertainties are given in % (1σ or 2σ confidence level): uncertainty associated with measurements (i.e., combination of statistical and systematic uncertainty), uncertainty due to nuclear data (noted ND) and total uncertainty (i.e., combination in quadrature of the first two terms). 235 U (n,f) Nuclide Yield Meas unc. % 2σ 95 Zr 6.42 4.1 97 Zr 5.84 4.4 99 Mo 6.18 4.0 103 Ru 2.97 4.2 105 Rh 0.92 5.5 131 I 2.86 4.3 132 Te 4.23 4.0 140 Ba 6.15 4.2 141 Ce 5.92 4.0 143 Ce 5.81 4.2 147 Nd 2.11 4.2

Thermal spectrum ND unc. 1σ 0.3 0.1 0.3 1.3 3.2 1.0 3.4 0.9 0.4 0.9 4.2

Total unc. 1σ 2.1 2.2 2.0 2.5 4.2 2.4 4.0 2.3 2.0 2.3 4.7

JEFF-3.1 Rel. diff. % % 6.50(1) -1.26 5.99(1.4) -2.62 6.13(1.5) +0.77 3.10(2.7) -5.11 0.95(1.1) -2.69 2.88(1.1) -0.63 4.28(1) -1.07 6.31(1.5) -2.60 5.86(2.5) +1.20 5.94(1.4) -2.25 2.22(1.8) -5.05

ENDF/B-VII.0 Rel. diff. % % 6.50(1.4) -1.29 5.98(2) -2.46 6.11(1.4) +1.15 3.03(1.4) -2.05 0.96(2) -4.80 2.89(1) -1.07 4.29(1.4) -1.53 6.21(1) -1.05 5.85(1) +1.23 5.96(1.4) -2.54 2.25(1.4) -5.43

TABLE VII: Results of fission product yield measurements for 239 Pu in a thermal neutron flux compared with JEFF-3.1 [13] and ENDF/B-VII.0 [14]. Uncertainties are given in % (1σ or 2σ confidence level): uncertainty associated with measurements (i.e., combination of statistical and systematic uncertainty), uncertainty due to nuclear data (noted ND) and total uncertainty (i.e., combination in quadrature of the first two terms). 239

Pu (n,f) Nuclide Yield Meas unc. % 2σ 95 Zr 4.81 4.6 97 Zr 5.31 3.7 99 Mo 6.54 3.5 103 Ru 7.08 4.0 131 I 4.06 4.0 132 Te 5.25 3.8 140 Ba 5.43 4.5 141 Ce 5.54 3.7 143 Ce 4.39 4.0 147 Nd 2.02 4.0

Thermal spectrum ND unc. 1σ 0.3 0.1 0.3 1.3 1.0 3.4 0.9 0.4 0.9 4.2

Total unc. 1σ 2.3 1.8 1.8 2.4 2.2 3.9 2.4 1.9 2.2 4.7

JEFF-3.1 Rel. diff. % % 4.95(2.0) -2.80 5.25(1.4) +1.18 6.18(0.9) +5.75 6.95(1.2) +1.90 3.72(2.1) +9.02 5.09(1.8) +3.05 5.32(1.1) +2.03 5.21(1.4) +6.43 4.48(1.1) -1.91 2.02(1.9) -0.22

the ratio to the JEFF-3.1 values and 0.81 for the ratio to ENDF/B-VII.0 values, confirming the very good agreement discussed above. For the 239 Pu target, we obtain larger χ2 values, i.e. 2.5 for the ratio to JEFF-3.1 and 1.4 for the ratio to ENDF/B-VII.0. While these values are larger that the 235 U ones, they also reflect the poorer knowledge of 239 Pu thermal neutron fission yields, as exhibited by the large dispersion between the average χ2 values per degree of freedom calculated using both evaluations. Yet, the average χ2 value of 1.4 is not so bad, considering the empirical evidence discussed above, for an isotope independent relative uncertainty of the order of 1 to 2.5 %.

ENDF/B-VII.0 Rel. diff. % % 4.82(1.4) -0.16 5.33(2.0) -0.44 6.21(1.4) +5.29 6.99(2.0) +1.22 3.86(1.0) +5.28 5.14(2.0) +2.15 5.35(1.4) +1.41 5.25(2.0) +5.59 4.41(1.4) -0.53 2.00(2.8) +0.85

C.

Fission spectrum 1.

233

U

Three experiments have been performed at the Prospero-U critical assembly. Two fission chambers were used with deposits respectively of 5.10 ± 0.05 μg and 6.40 ± 0.06 μg. The fission chamber-target setups were encapsulated with cadmium and located 166 mm away from the assembly axis. Simulations with the Tripoli4 [9] and MCNP 4C [15] codes show that at this location, taking into account the 0.8 mm-thick cadmium encapsulation and using the 233 U fission cross section from ENDF/B-VI, the average energy of the neutrons leading to fission, < Efiss >, in the 233 U sample is  0.77 MeV.

2974

Ratio exp/eval

Fission Products Yields of

233

U,

235

U, . . .

NUCLEAR DATA SHEETS

235U Thermal

1.1

TABLE VIII: Fission product yields for 233 U in a fission neutron spectrum and for 14.7 MeV incident neutrons. Relative uncertainties (1σ or 2σ confidence level) are quoted in %: uncertainty associated with measurements (i.e., combination of statistical and systematic uncertainty), uncertainty due to nuclear data (noted ND) and total uncertainty (i.e., combination in quadrature of the first two terms).

1.05 1

0.95 0.9

Ratio exp/eval

100

110

120

130

140 Mass Number

130

140 Mass Number

233

U Fission spectrum 14.7 MeV (n,f) < Efiss > 0.77 MeV Nuc- Yield Meas ND Tot. Yield Meas ND Tot. lide unc. unc. unc. unc. unc. unc. % 2σ 1σ 1σ 2σ 1σ 1σ 95 Zr 6.34 5.1 0.3 2.6 5.08 6.0 0.3 3.0 97 Zr 5.46 4.1 0.1 2.1 4.80 6.2 0.1 3.1 99 Mo 4.94 4.0 0.3 2.0 4.43 4.1 0.3 2.1 103 Ru 1.65 5.4 1.3 3.0 2.75 3.8 1.3 3.1 131 I 3.84 4.1 1.0 2.3 4.66 3.8 1.0 2.3 132 Te 4.62 4.0 3.4 4.0 3.14 3.8 3.4 3.9 133 I 5.54 4.2 3.3 3.9 135 Xe 5.44 8.5 3.3 4.2 3.48 7.5 3.3 5.0 136 Cs 0.11 7.8 1.5 4.2 0.83 6.5 1.5 4.9 137 Cs 6.17 12.5 0.2 6.3 140 Ba 5.96 4.2 0.9 2.3 4.30 6.2 0.9 3.1 141 Ce 6.31 4.1 0.4 2.1 4.25 6.6 0.4 3.3 143 Ce 5.28 4.2 0.9 2.3 3.11 6.0 0.9 3.1 144 Ce 4.53 5.3 1.7 3.2 2.44 8.5 1.7 4.5 147 Nd 1.64 5.4 4.2 5.0 1.23 7.0 4.2 7.2

239Pu Thermal

1.1

1.05 1

0.95 0.9 100

110

120

FIG. 5: Yields for 235 U and 239 Pu with thermal neutrons: ratio to evaluated yields from JEFF-3.1 (full symbols) and ENDF/B-VII.0 (open symbols). Error bars include the total uncertainties of the measurements and the uncertainties from the evaluations. A great similarity between 235 U and 239 Pu is observed in the variations of the ratio from one fission product to the other. This most likely reflects errors in nuclear data and/or the gamma detection efficiency curve.

The integrated flux experienced by the targets during each irradiation was on the order of 1.6×1015 neutrons. In these experimental conditions the neutron flux gradient induces a 4 % readjustment of the flux between the fission chamber deposit and the target. The results of the three experiments are consistent with each other and the final values obtained are given in Tables VIII, and compared with evaluated values from the JEFF-3.1 [13] and ENDF/BVII [14] libraries in Fig. 6. For convenience, the 14-MeV results are also included in this and several subsequent tables.

2.

235

J. Laurec et al.

U

235 U was used to design the experimental procedure for the whole measurement program. As a consequence many irradiations and measurements have been performed for this actinide. Five fission chambers were used in more than eight irradiations. The precise number of irradiations as well as of that of gamma spectrometry measurements is not available today. The irradiations of the fission chamber experiments at Prospero were performed outside the assembly. Most of the irradiations at Caliban were performed within the central cavity. The relevant neutron spectral shapes are given at figure 3. The deposits of two of the fission chambers were extracted subsequently to the experiments to perform alpha and mass spectrometry measurements in order to crosscheck the deposited mass of 235 U. The effect of fission neutron spectra on fission yields

has been studied on 235 U by comparing the results from irradiations in the Caliban central cavity and irradiations outside the Prospero and Caliban assemblies. As mentioned earlier, the neutron energy spectra at Caliban and outside Prospero are largely different. The mean neutron energy at Caliban is ∼ 1.35 MeV, both within the cavity and outside the assembly, whereas outside the Prospero assembly it is close to 0.75 MeV. However, as reported in Ref. [6], no significant effect on the fission product yields could be seen within the uncertainties. As a consequence the results from the different neutron spectra have been combined. Based on the characteristics of the irradiations (power, duration and position of the target setups, target mass etc.), we estimate the average energy of the neutrons leading to fission (called < Efiss > in this article) corresponding to the fission chamber experiments on 235 U as being on the order of 1.25 MeV. A specific experiment has been performed for determining the product yields of 155 Eu and 156 Eu. The low value of the production yields for these two fission products (on the order of 10−4 ) required larger masses of fissile material and longer irradiation periods. Since a direct measurement of the activity on the targets was not possible, they have been put into solution and the cerium, neodymium and europium rare earths have been chemically separated. The number of fissions induced in the target was obtained from the activity of 141 Ce, 143 Ce, 144 Ce, and 147 Nd using the fission product yields previously obtained for these nuclides in the fission chamber

2975

Fission Products Yields of Ratio exp/eval

1.2

233

U,

235

U, . . .

233U Fission spectrum

TABLE IX: Fission product yields of Eu isotopes for 235 U irradiated in a fission neutron spectrum, with different irradiation conditions. Relative uncertainties (1σ or 2σ confidence level) are quoted in %: uncertainty associated with measurements (i.e., combination of statistical and systematic uncertainty), uncertainty due to nuclear data (noted ND) and total uncertainty (i.e., combination in quadrature of the first two terms).

1.1 1

0.9 0.8

100

110

Ratio exp/eval

1.2

120

130

140 Mass Number

235

U(n,f) Nuclide and conditions

235U Fission spectrum

1.1

155

Eu Cd shielding outside assembly 155 Eu B+Cd shielding outside assembly . 155 Eu cavity 156 Eu Cd shielding outside assembly 156 Eu B+Cd shielding outside assembly 156 Eu cavity

1

0.9 0.8

100

110

Ratio exp/eval

1.2

120

130

140 Mass Number

239Pu Fission spectrum

1.1 1

0.9 0.8

100

110

233

120

235

130

J. Laurec et al.

NUCLEAR DATA SHEETS

140 Mass Number

239

FIG. 6: Yields for U, U and Pu with fast neutrons: ratio to evaluated yields from JEFF-3.1 (full symbols) and ENDF/B-VII.0 (open symbols). Error bars include the total uncertainties of the measurements and the uncertainties from the evaluations.

experiments. The 155,156 Eu yields reported here are thus measurements relative to the absolute measurements of 141 Ce, 143 Ce, 144 Ce and 147 Nd yields. In order to study the possible influence of the neutron spectrum on the 155 Eu and 156 Eu fission yields, three samples have been irradiated in the Prospero-U critical assembly. Each sample was placed in a cylindrical aluminum container. • The first container was placed in the Prospero cavity. It contained 400 mg of uranium oxide. • A second container with a 0.8 mm thick cadmium encapsulation was placed outside the assembly. It contained 2 grammes of uranium oxide. • A third container, identical to the second one, had an extra 1 cm-thick encapsulation made of boron carbide. This protection is equivalent to 7 mm boron and results in an additional neutron attenuation of a factor of 27 at 10 eV and 2.2 at 100 eV. To collect all the fission products, the uranium oxide samples were wrapped into aluminum foils which were dissolved together with the uranium oxide. Irradiation duration was 2×105 seconds for these specific experiments.

< Efiss > (MeV) 0.7

Fission spectrum Yield Meas ND Total unc. unc. unc. % 2σ 1σ 1σ 3.97×10−2 8.5 1.3 4.4

0.8

3.71×10−2

8.5

1.3

4.4

1.2

3.98×10−2

7.5

1.3

4.0

0.7

1.77×10−2

8.5 11.8

12.5

0.8

1.67×10−2

8.5 11.8

12.5

1.2

1.86×10−2

7.5 11.8

12.4

The results are reported in Table IX for the three different irradiation conditions. Mean neutron energy leading to fission obtained from Monte Carlo simulations are also reported. Although the results remain consistent within the error bars, there is a trend in the central cavity spectrum yielding higher values compared with exterior spectra for both europium isotopes. Another relative measurement involving two samples, of 52 and 57 mg, of uranium oxide was performed at the Prospero-U assembly in 1980 [5]. The targets were placed in containers made of high purity aluminum, themselves wrapped into 1 mm-thick cadmium foils. During the irradiations the samples were located outside the assembly. The number of fissions was measured by counting several fission products easily measurable by gamma spectrometry and for which the fission yields have been previously measured in the fission chamber experiments. The fission product mass chains used for this purpose were 99, 103, 132, 140, 141 and 143. On one of the two targets a chemical isolation of Ce has been performed. This experiment allowed us to obtain independent data points for the 95 Zr, 144 Ce and 147 Nd fission product yields. Since the irradiation points were located outside the Prospero assembly, the average energy leading to fission in the sample was ∼ 0.7 MeV. The values of the fission product yields obtained in these three series of experiments and the associated uncertainties are given in Table X. For 95 Zr, 144 Ce and 147 Nd, two results are given which correspond respectively to the fission chamber experiments and the relative

2976

Fission Products Yields of

233

U,

235

U, . . .

NUCLEAR DATA SHEETS

TABLE X: Fission product yields for 235 U in a fission neutron spectrum and for 14.7 MeV incident neutrons. Relative uncertainties (1σ or 2σ confidence level) are quoted in %: uncertainty associated with measurements (i.e., combination of statistical and systematic uncertainty), uncertainty due to nuclear data (noted ND) and total uncertainty (i.e., combination in quadrature of the first two terms). The results from relative measurements are denoted by an asterisk. 235 U (n,f)

Nuclide 95

Zr Zr∗ 97 Zr 99 Mo 103 Ru 106 Ru 127 Sb . 131 I 132 Te 133 I 135 Xe 136 Cs 137 Cs 140 Ba 141 Ce 143 Ce 144 Ce 144 Ce∗ 147 Nd 147 Nd∗ 95

Fission spectrum 14.7 MeV < Efiss > 1.25 MeV * < Efiss > 0.7 MeV Yield Meas ND Total Yield Meas ND Total unc. unc. unc. unc. unc. unc. % 2σ 1σ 1σ % 2σ 1σ 1σ 6.32 4.2 0.3 2.1 5.28 4.3 0.3 2.2 6.23 4.5 0.3 2.3 5.82 3.8 0.1 1.9 4.98 5.5 0.1 2.8 6.27 3.4 0.3 1.7 5.08 4.3 0.3 2.2 3.18 4.1 1.3 2.4 3.02 4.6 1.3 2.7 0.56 10.5 2.3 5.7 - 1.73 8.0 4.1 5.7 3.34 3.6 1.0 2.1 4.35 5.5 1.0 2.5 4.76 3.5 3.4 3.9 4.14 4.4 3.4 4.1 5.90 2.9 3.3 4.5 5.43 5.4 3.3 4.3 - 5.01 6.9 3.3 4.8 - 0.214 8.5 1.5 5.0 6.11 9.0 0.2 4.5 5.84 4.4 0.9 2.4 4.52 4.7 0.9 2.5 5.85 4.4 0.4 2.2 4.52 4.5 0.4 2.3 5.31 3.7 0.9 2.1 3.68 4.6 0.9 2.5 4.97 5.0 1.7 3.0 3.16 1.5 1.7 1.9 5.08 5.0 1.7 3.0 2.04 5.0 4.2 4.9 1.51 5.3 4.2 5.0 2.12 5.0 4.2 4.9 -

measurement described above.

3.

239

Pu

The fission chamber experiments on 239 Pu have been conducted simultaneously with those on 235 U. The experimental conditions are thus identical with however additional difficulties due to target conditioning. We believe more than eight irradiations were conducted although most of the archive is lost. The average energy of the neutrons leading to fission in the target has been calculated from simulations using the Tripoli 4 and MCNP 4C codes for the different critical assemblies and target positions and using the fission cross section from ENDF/B-VI. The cadmium encapsulation was taken into account for the irradiations outside the assemblies. We obtain the following results: • Caliban, in cavity < Efiss >1.4 MeV

and

outside

assembly:

• Prospero, outside assembly: < Efiss > 0.8 MeV

J. Laurec et al.

Based on the characteristics of the different irradiations and the target masses we estimate the average < Efiss > corresponding to the final results as about 1.25 MeV. As for 235 U, a relative measurement of the 95 Zr, 144 Ce and 147 Nd fission products has been performed in 1980 at the Prospero-U critical assembly [5]. Since the gamma activity of 144 Ce was too low with respect to that of 239 Pu and the other fission products, its fission yield could not be determined in the fission chamber experiments. Dedicated experiments involving larger masses of 239 Pu and longer irradiation times were thus required. Two targets, respectively of 18 mg and 27 mg of plutonium oxide were irradiated in the Prospero critical assembly during 105 seconds. The number of fissions have been obtained from the activities of several fission products easily measurable by gamma spectrometry and for which the fission product yield had been previously determined. After the few days during which the gamma spectrometry measurements were performed, one of the targets was put into solution and a chemical isolation of cerium was performed. The second target was regularly measured during several months. The result on the 144 Ce yield given in Table XI corresponds to the weighted average of those obtained by the two methods. The fission yield of 106 Ru also given in Table XI was obtained from the same study. For these results, the average energy of the neutrons leading to fission is ∼ 0.8 MeV. The results from the two series of experiments are given in Table XI. The utilization in Ref. [37] of the results given in Table XI for fission neutrons seems to indicate that our 239 Pu fission spectrum measurements are systematically lower than other comparable data. On Fig. 6, a comparison is performed between our measured yields in fission spectrum and comparable evaluated data from the JEFF3.1 [13] and ENDF/B-VII.0 [14] libraries. While for 233 U and 235 U the average of the ratios are fairly close to unity (underestimated by about 1 to 1.5%), our measured 239 Pu yields underestimate the evaluated yields by about 4.5 % on average (which is larger than the uncertainties quoted in Table. XI). Relating these trends to the trends exhibited by our thermal data (about -1% for 235 U and about +2.5% for 239 Pu as shown in Sec. IV B) confirms the high consistency of our 235 U measurements, but points to systematic errors associated with our 239 Pu measurements that are likely higher that the values quoted in Table XI. To further investigate the unaccounted for bias that seems to be affecting our 239 Pu data, the ratios of thermal to fission spectrum yields for 239 Pu were constructed in the 100 and 140 mass regions. In these mass regions, located close to the peaks of the fission product yields distributions, fission product yields are expected to exhibit minimal neutron energy dependence. With our data, this ratio is close to 0.91, whereas it is of the order of 0.975 when calculated with the data from the England-Rider evaluation [38] or with the Wahl systematics on fission product yields [39]. Taking into account the 2.5% over-

2977

Fission Products Yields of

233

U,

235

U, . . .

TABLE XI: Fission product yields for 239 Pu in a fission neutron spectrum and for 14.7 MeV incident neutrons. Relative uncertainties (1σ or 2σ confidence level) are quoted in %: uncertainty associated with measurements (i.e., combination of statistical and systematic uncertainty), uncertainty due to nuclear data (noted ND) and total uncertainty (i.e., combination in quadrature of the first two terms). The results from relative measurements are denoted by an asterisk. 239 Pu (n,f)

Nuclide 95

Zr Zr∗ 97 Zr 99 Mo 103 Ru 105 Rh 106 Ru∗ 127 Sb 131 I 132 Te 133 I 136 Cs 140 Ba 141 Ce 143 Ce 144 Ce 144 Ce∗ 147 Nd 147 Nd∗ 95

J. Laurec et al.

NUCLEAR DATA SHEETS

Fission spectrum 14.7 MeV < Efiss > 1.25 MeV * < Efiss > 0.8 MeV Yield Meas ND Total Yield Meas ND Total unc. unc. unc. unc. unc. unc. % 2σ 1σ 1σ % 2σ 1σ 1σ 4.47 4.1 0.3 2.1 3.92 4.4 0.3 2.2 4.44 4.0 0.3 2.0 4.97 4.2 0.1 2.1 4.36 5.0 0.1 2.5 5.98 3.5 0.3 1.8 5.07 4.1 0.3 2.1 6.45 3.7 1.3 2.3 5.12 4.1 1.3 2.5 5.16 5.4 3.1 4.1 4.25 5.1 3.1 4.0 3.99 7.0 2.3 4.2 - 2.13 4.9 4.1 4.8 3.65 5.2 1.0 2.8 4.58 5.3 1.0 2.8 4.87 3.8 3.4 3.9 3.15 4.4 3.4 4.0 5.92 6.5 3.3 4.7 4.25 9.5 3.3 5.8 0.11 7.0 1.5 3.8 0.79 4.9 1.5 2.9 4.88 5.0 0.9 2.6 3.53 4.3 0.9 2.3 4.86 4.8 0.4 2.4 3.69 4.6 0.4 2.3 4.00 4.1 0.9 2.3 2.87 4.5 0.9 2.4 - 2.60 7.8 1.7 4.3 3.59 4.5 1.7 2.8 1.86 4.3 4.2 4.8 1.53 4.7 4.2 4.8 1.86 4.5 4.2 4.7 -

D.

14 MeV neutrons 1.

233

U

Two experiments have been performed using the Lancelot neutron generator. The fission chambers are the same which were used during the fission neutron experiments. The experimental conditions were described previously. Gamma ray spectra analyses were more difficult due to the low activity of most of the measured fission products and the large activity of the daughter nuclides of 232 U present in the fissile material. However, the results of the two experiments are in good agreement. The results obtained are given in Table VIII and compared with evaluated values from the JEFF-3.1 [13] and ENDF/BVII [14] libraries in Fig. 7.

2.

235

U

Four irradiations were performed with different parameters: targets, fission chambers, distance between sample and tritium target. The results from the different experiments are in good agreement. Final results are given in Table X and compared with evaluated values from the JEFF-3.1 [13] and ENDF/BVII [14] libraries in Fig. 7.

3.

238

U

For the two experiments realized on this element two fission chambers have been used. The results are listed in Table XII and compared with evaluated values from the JEFF-3.1 [13] and ENDF/BVII [14] libraries in Fig. 7.

4.

estimation of our 239 Pu thermal spectrum values documented in Sec. IV B leaves a difference of the order of 4%, of which 2 to 2.5% can be accounted for by the errors quoted in Table XI, leaving an unaccounted for downward bias of the order or 2% in our 239 Pu fission product yields measurements in fission spectrum. We have deeply investigated possible sources of biases or systematic errors that could have been neglected in the original analysis of Ref. [6], and after performing many cross-checks and interviewing the original authors, we could not find any such overlooked systematic error. Nevertheless, in the specific case of 147 Nd, comparisons with thermal 235 U and 239 Pu (Table VI) suggests that the large underestimation of the 147 Nd yield might be due, to some extent, to the values of the gamma intensities and half-lives used to transform the corrected Ge(Li) peak areas to number of atoms.

239

Pu

Three experiments have been performed for which the results are in good agreement. The final results are listed in Table XI and compared with evaluated values from the JEFF-3.1 [13] and ENDF/BVII [14] libraries in Fig. 7.

V.

CONCLUSION

We report the results of a large, multi-years, fission yields measurement program, performed at CEA DAM in the seventies. More than one hundred fission yields were measured for the neutron induced fission of 233 U, 235 U, 238 U and 239 Pu in thermal, fission (fast critical assembly) and 14.7 MeV neutron spectra, for up to 19 fission products. Absolute yields were measured using gamma spectroscopy to count the production of the fission products of interest in irradiated samples, through well known gamma lines, and normalized to a number fission counted

2978

Fission Products Yields of Ratio exp/eval

1.2

233

U,

235

U, . . .

energy applications. The measurement of 235 U (and to a lesser extent 239 Pu) fission yields in a thermal spec-

233U 14.7 MeV

1

TABLE XII: Fission product yields for 238 U for 14.7 MeV incident neutrons. Relative uncertainties (1σ or 2σ confidence level) are quoted in %: uncertainty associated with measurements (i.e., combination of statistical and systematic uncertainty), uncertainty due to nuclear data (noted ND) and total uncertainty (i.e., combination in quadrature of the first two terms).

0.8 100

110

Ratio exp/eval

1.2

120

130

140 Mass Number

235U 14.7 MeV

238

U(n,f) 14.7 MeV Nuclide Yield Meas ND Total unc. unc. unc. % 2σ 1σ 1σ 95 Zr 4.92 4.7 0.3 2.4 97 Zr 5.18 5.3 0.1 2.7 99 Mo 5.79 4.4 0.3 2.2 103 Ru 4.64 4.5 1.3 2.6 105 Rh 3.36 6.5 3.1 4.5 127 Sb 1.35 5.7 4.1 5.0 131 I 4.08 4.8 1.0 2.6 132 Te 4.72 4.4 3.4 4.1 133 I 5.50 6.5 3.3 4.7 135 Xe 5.74 5.5 3.3 4.3 140 Ba 4.56 4.7 0.9 2.5 141 Ce 4.45 4.5 0.4 2.3 143 Ce 3.86 5.0 0.9 2.7 144 Ce 3.73 7.5 1.7 4.1 147 Nd 1.94 5.2 4.2 4.9

1

0.8 100

110

Ratio exp/eval

1.2

120

130

140 Mass Number

130

140 Mass Number

130

140 Mass Number

238U 14.7 MeV

1

0.8 100

110

1.2

Ratio exp/eval

J. Laurec et al.

NUCLEAR DATA SHEETS

120 239Pu 14.7 MeV

1

0.8 100

110

120

FIG. 7: Yields for 233 U, 235 U, 238 U and 239 Pu with fast neutrons: ratio to evaluated yields from JEFF-3.1 (full symbols) and ENDF/B-VII.0 (open symbols). Error bars include the total uncertainties of the measurements and the uncertainties from the evaluations.

with fission chambers. The samples were irradiated at the Saclay EL3 reactor for thermal neutrons, at the Prospero and Caliban fast critical assemblies in CEA Valduc for fission neutrons and at the Lancelot accelerator in Valduc for 14.7 MeV neutrons. Characterization of the mass and isotopic abundances of the actinide samples, as well as of the gamma and fission chamber detection efficiencies were performed with great care. The statistical and systematic uncertainties associated with each potential source of error were carefully assessed and quantified. The results, which were previously only available in a CEA report [6], are now updated using the most up to date nuclear decay data, including updated uncertainties. These measurements constitute a new, independent, internally consistent data set, to be included in the evaluation of the fission yield nuclear data used for nuclear

trum, allows this whole data set to establish contact with the extensively measured and well known thermal 235 U body of fission yield data. Comparisons between our measured thermal yields and current evaluated ones [13, 14] show a very high level of consistency between our thermal 235 U measurements and the corresponding evaluated yield data. While the agreement between our measured 239 Pu thermal neutron fission yields and evaluations is not as spectacular as that of 235 U, it allows us to empirically quantify the different types of uncertainties (isotope dependent or isotope-independent) associated with our measurements. Similar levels of agreement can also be achieved for other irradiation spectra, as illustrated by the comparisons with evaluated values displayed on Figs. 6, and 7 in fission and 14.7-MeV neutron fields, respectively.

Acknowledgments

We thank the members of the Expert Panel [40] for encouraging us to publish the present work : T. Baisden, J. Ferguson, D. Gilliam, R. Jeanloz, C. McMillan, D. Robertson, P. Thompson, C. Verdon, C.W. Wilkerson, and P.G. Young. We also thank J. Kenneally and M.B. Chadwick for very fruitful discussions and comments.

2979

Fission Products Yields of

233

U,

235

U, . . .

NUCLEAR DATA SHEETS

[1] International Criticality Safety Benchmark Evaluation Project Handbook 2009, entry HEU-MET-FAST-080, (Nuclear Energy Agency OECD, Paris, 2009). [2] D. Gilliam, et al., InterLaboratory Reaction Rate Program, 11th Progress Report, Hanford Engineering Development Laboratory report HEDL-TME 77-34, UC-79 b,d (1977). [3] J. Laurec, Internal CEA DAM report (1976). [4] J. Laurec, Internal CEA DAM report (1978). [5] J. Laurec, Internal CEA DAM report (1981). [6] J. Laurec, A. Adam, T. De Bruyne, D´etermination des rendements de fissions induites par un spectre de neutrons de fission et des neutrons de 14.7 MeV dans 233 U, 235 U, 238 U, 239 Pu, report CEA-R-51470 (CEA, Saclay, 1981). [7] J. Robert, J. Hainzelin, V. Raievski, Pile EL 3, CEA Report no 1161, Saclay 1959. [8] International Group on Research Reactors internet site, http://www.igorr.com, pages updated in 2007. [9] J. P. Both, A. Mazzolo, O. Petit, Y. Peneliau, B. Roesslinger, User Manual for version 4.3 of the TRIPOLI-4 Monte Carlo method particle transport computer code, CEA-Report: CEA-R-6044, DTI, CEA/Saclay, France, 2003. [10] C. Bergey, CEA-DAM internal report (1976). [11] J.P. Perrine, Contribution ` a l’´etude et au d´eveloppement de l’´electropulv´erisation : application a ` la fabrication de d´epots uniformes, Commissariat ` a l’Energie Atomique report CEA-R-4780 (1976). [12] R.J. Gehrke, R.G. Helmer, R.C. Greenwood, Precise Relative γ-Ray Intensities for Calibration of Ge Semiconductor Detectors, Nucl. Inst. and Methods. 147, 405 (1977). [13] M.A. Kellett, O. Bersillon, R.W. Mills, The JEFF3.1/3.1.1 radioactive decay data and fission yields sublibraries, JEFF report 20, (NEA/OECD, Paris 2009). [14] M.B. Chadwick et al., ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology, Nuclear Data Sheets 107, 2931 (2006). [15] MCNP A General Monte Carlo N-Particle Transport Code version 4C, J.F. Briestmeister ed., LA-13709-M, Los Alamos (2000). [16] T. W. Burrows, Nuclear Data Sheets Update for A = 95, Nuclear Data Sheets 68,635 (1993). [17] N. Nica, Nuclear Data Sheets for A = 97, Nuclear Data Sheets 111, 525 (2010). [18] L. K. Peker, Nuclear Data Sheets Update for A = 99, Nuclear Data Sheets 73,1 (1994). [19] D. De Frenne, Nuclear Data Sheets for A = 103, Nuclear Data Sheets 110, 2081 (2009). [20] D. De Frenne, E. Jacobs, Nuclear Data Sheets for A = 105, Nuclear Data Sheets 105, 775 (2005). [21] D. De Frenne, A. Negret, Nuclear Data Sheets for A = 106, Nuclear Data Sheets 109, 943 (2008).

J. Laurec et al.

[22] K. Kitao, M. Oshima, Nuclear Data Sheets Update for A = 127, Nuclear Data Sheets 77, 1 (1996). [23] Shaheen Rab, Nuclear Data Sheets Update for A = 133, Nuclear Data Sheets 75, 491 (1995). [24] Yu. Khazov, I. Mitopolsky, A. Rodionov, Nuclear Data Sheets for A = 131, Nuclear Data Sheets 107, 2715 (2006). [25] Yu. Khazov, A. Rodionov, S. Sakharov, Balraj Singh, Nuclear Data Sheets for A = 132, Nuclear Data Sheets 104, 497 (2005). [26] Balraj Singh, A. Rodionov, Yu. Khazov, Nuclear Data Sheets for A = 135, Nuclear Data Sheets 109, 517 (2008). [27] A.A. Sonzogni, Nuclear Data Sheets for A = 136 , Nuclear Data Sheets 95, 837 (2002). [28] E. Browne, J. K. Tuli, Nuclear Data Sheets for A = 137, Nuclear Data Sheets 108, 2173 (2007). [29] N. Nica, Nuclear Data Sheets for A = 140, Nuclear Data Sheets 108, 1287 (2007). [30] J. K. Tuli, D. F. Winchell, Nuclear Data Sheets for A = 141, Nuclear Data Sheets 92, 277 (2001). [31] J. K. Tuli, Nuclear Data Sheets for A = 143, Nuclear Data Sheets 94, 605 (2001). [32] A.A. Sonzogni, Nuclear Data Sheets for A = 144, Nuclear Data Sheets 93, 599 (2001). [33] N. Nica, Nuclear Data Sheets for A = 147, Nuclear Data Sheets 110, 749 (2009). [34] C. W. Reich, Nuclear Data Sheets for A = 155, Nuclear Data Sheets 104, 1 (2005). [35] C. W. Reich, Nuclear Data Sheets for A = 156, Nuclear Data Sheets 99, 753 (2003). [36] J.A. Grundl, D.M. Gilliam, N.D. Dudey, R.J. Popek, Measurement of Absolute Fission Rates, Nuclear Technology Vol 25, No 2, 237 (1975). [37] M.B. Chadwick et al., Fission Product Yields from Fission Spectrum n+239 Pu for ENDF/B-VII.1, Nuclear Data Sheets 111, 2923 (2010). [38] T. R. England and B. F. Rider, Evaluation and Compilation of Fission Product Yields ENDF-349, LA-UR-943106, Los Alamos National Laboratory (1994). [39] A.C. Wahl, Systematics of Fission products Yields, Los Alamos National Laboratory report LA-13928, May (2002). [40] T. Baisden, E. Bauge, J. Ferguson, D. Gilliam, T. Granier, R. Jeanloz, C. McMillan, D. Robertson, P. Thompson, C. Verdon, C.W. Wilkerson, P, G, Young, Measurement / Evaluation Techniques and Nuclear Data Associated with Fission of 239 Pu by Fission Spectrum Neutrons, Los Alamos and Lawrence Livermore National Laboratory Reports, LAU-R-10-01598 and LLNL-TR42165 (2010).

2980