Measurement of Neutron Capture Cross Section of 237 Np from 0.02 to 100 eV Oleg SHCHERBAKOV 1; y , Kazuyoshi FURUTAKA 1 , Shoji NAKAMURA 1 , Hitoshi SAKANE 1 , Katsuhei KOBAYASHI 2 , Shuji YAMAMOTO 2 , Jun-Ichi HORI 2 and Hideo HARADA 1; 1 Tokai Works, Japan Nuclear Cycle Development Institute, Tokai-mura, Naka-gun, Ibaraki 319-1194 2 Research Reactor Institute, Kyoto University, Noda, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Received May 26, 2004 and accepted in revised form November 4, 2004) The neutron capture cross section of 237 Np has been measured relative to the 10 B(n;) 7 Li cross section by the time-of-flight method in the energy range from 0.02 to 100 eV. The 46 MeV electron linear accelerator at the Research Reactor Institute, Kyoto University, was used as a pulsed neutron source. The BGO scintillation detector was em- ployed in conjunction with the flash ADC-based data taking system for measurement and data accumulation. For the first time the capture cross section of 237 Np in resonance energy range was measured using the total energy gam- ma-ray detector. The results of present measurements have been compared with the evaluated capture cross sections of ENDF/B-VI and JENDL-3.3, as well as with the data measured by other authors. KEYWORDS: neptunium 237, neutron capture cross section, measurement, time-of-flight method, capture gamma rays, BGO detector, neutron resonances I. Introduction Neutron capture cross sections of long-lived fission prod- ucts and minor actinides have received much attention dur- ing last decade in the field of nuclear transmutation of radio- active waste. However, accurate measurement of these cross sections is a very difficult task due to high natural radioactiv- ity of samples under investigation. Neptunium-237 is one of the most important nucleus in the field of nuclear transmuta- tion, and also for the study of high burn-up behavior of nu- clear fuel. In the past, a few experimental groups have measured the neutron capture cross section of 237 Np in the resonance ener- gy range below 200 eV. 1–4) These measurements were per- formed by the neutron time-of-flight (TOF) method using an underground nuclear explosion 1) or an electron Linac 2–4) as a pulsed neutron source. Another important feature of all these experiments was a use of Moxon–Rae 1) and hydrogen-free C 6 F 6 2) or C 6 D 6 3,4) detectors which are characterized by a small efficiency for capture gamma-rays. In two cases, 2,3) a pulse-height weighting technique 5) has been applied to ex- clude a dependence of the detector efficiency on the possible variations of spectra and multiplicity of the prompt capture gamma-ray cascade from resonance to resonance. To reduce the high background caused by gamma rays of the radioac- tive neptunium sample, the pulse-height bias corresponding to 0.4–0.95 MeV was used. 2,4) Even with employment of this comparatively high bias, it was necessary to insert the 10- mm-thick lead shield 2) between the sample and detectors or to arrange coincidences between the detectors, 4) in order to obtain a good foreground-to-background ratio. In the present experiment, another approach was applied by making use of a total energy absorption detector. Capture events registered by such a detector give rise to pulse heights proportional to B n þE n , where B n is the neutron separation energy of 238 Np (5.488 MeV) which are constant with a good accuracy because B n E n . Besides the constant detection ef- ficiency for the capture event, a total absorption detector en- ables to separate effectively capture and natural background gamma rays because the latter have the energy much lower than B n þE n . The 8.54 liter 4% bismuth germanate (BGO) scintillation detector and the associated flash analog-to-digital converter (ADC)-based data taking system developed at the Japan Nuclear Cycle Development Institute (JNC) have been previously reported elsewhere 6) and are not described here in detail. With this detector, the measurements of the neutron capture cross section of 237 Np from 0.02 to 100 eV have been carried out by the neutron TOF method using the 46 MeV electron Linac at Research Reactor Institute, Kyoto Univer- sity (KURRI). 7) II. Measurement Technique 1. Experimental Set-up The experimental set-up is shown in Fig. 1. The measure- ments were performed under the following Linac operating conditions: the electron energy, 30 MeV; pulse repetition rate, 30–100 Hz; pulse width, 100 ns–3 ms; and average elec- tron current, 40–60 mA. The accelerated electrons struck a water-cooled Ta target, producing fast neutrons having the evaporation spectrum with an average energy of 0:7 MeV. Slow neutrons were produced by moderating fast neutrons in a 10 cm thick water moderator coupled with a neutron-pro- ducing target. To decrease the overloading effect caused in the detector phototubes by an intensive flash of gamma rays from the target, two kinds of target-moderator geometries have been tested. In the first case, a 10-cm-thick lead shadow bar was placed on the neutron beam axis close to the neu- tron-producing target in such a way that the sample could not see the target but only the moderator. In the second case, without the shadow bar, it was found that 30–40 mm dis- y On leave from Petersburg Nuclear Physics Institute, Russia. Corresponding author, Tel. +81-29-282-1111, Fax. +81-29-282- 1517, E-mail: [email protected]Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 42, No. 2, p. 135–144 (February 2005) 135 ORIGINAL PAPER
Transcript
Measurement of Neutron Capture Cross Section of 237Np from 0.02 to 100 eV
1Tokai Works, Japan Nuclear Cycle Development Institute, Tokai-mura, Naka-gun, Ibaraki 319-11942Research Reactor Institute, Kyoto University, Noda, Kumatori-cho, Sennan-gun, Osaka 590-0494
(Received May 26, 2004 and accepted in revised form November 4, 2004)
The neutron capture cross section of 237Np has been measured relative to the 10B(n; �)7Li� cross section by thetime-of-flight method in the energy range from 0.02 to 100 eV. The 46MeV electron linear accelerator at the ResearchReactor Institute, Kyoto University, was used as a pulsed neutron source. The BGO scintillation detector was em-ployed in conjunction with the flash ADC-based data taking system for measurement and data accumulation. Forthe first time the capture cross section of 237Np in resonance energy range was measured using the total energy gam-ma-ray detector. The results of present measurements have been compared with the evaluated capture cross sections ofENDF/B-VI and JENDL-3.3, as well as with the data measured by other authors.
Neutron capture cross sections of long-lived fission prod-ucts and minor actinides have received much attention dur-ing last decade in the field of nuclear transmutation of radio-active waste. However, accurate measurement of these crosssections is a very difficult task due to high natural radioactiv-ity of samples under investigation. Neptunium-237 is one ofthe most important nucleus in the field of nuclear transmuta-tion, and also for the study of high burn-up behavior of nu-clear fuel.
In the past, a few experimental groups have measured theneutron capture cross section of 237Np in the resonance ener-gy range below �200 eV.1–4) These measurements were per-formed by the neutron time-of-flight (TOF) method using anunderground nuclear explosion1) or an electron Linac2–4) as apulsed neutron source. Another important feature of all theseexperiments was a use of Moxon–Rae1) and hydrogen-freeC6F6
2) or C6D63,4) detectors which are characterized by a
small efficiency for capture gamma-rays. In two cases,2,3) apulse-height weighting technique5) has been applied to ex-clude a dependence of the detector efficiency on the possiblevariations of spectra and multiplicity of the prompt capturegamma-ray cascade from resonance to resonance. To reducethe high background caused by gamma rays of the radioac-tive neptunium sample, the pulse-height bias correspondingto 0.4–0.95MeV was used.2,4) Even with employment of thiscomparatively high bias, it was necessary to insert the 10-mm-thick lead shield2) between the sample and detectorsor to arrange coincidences between the detectors,4) in orderto obtain a good foreground-to-background ratio.
In the present experiment, another approach was appliedby making use of a total energy absorption detector. Captureevents registered by such a detector give rise to pulse heights
proportional to BnþEn, where Bn is the neutron separationenergy of 238Np (5.488MeV) which are constant with a goodaccuracy because Bn�En. Besides the constant detection ef-ficiency for the capture event, a total absorption detector en-ables to separate effectively capture and natural backgroundgamma rays because the latter have the energy much lowerthan BnþEn.
The 8.54 liter 4� bismuth germanate (BGO) scintillationdetector and the associated flash analog-to-digital converter(ADC)-based data taking system developed at the JapanNuclear Cycle Development Institute (JNC) have beenpreviously reported elsewhere6) and are not described herein detail. With this detector, the measurements of the neutroncapture cross section of 237Np from 0.02 to 100 eV have beencarried out by the neutron TOF method using the 46MeVelectron Linac at Research Reactor Institute, Kyoto Univer-sity (KURRI).7)
II. Measurement Technique
1. Experimental Set-upThe experimental set-up is shown in Fig. 1. The measure-
ments were performed under the following Linac operatingconditions: the electron energy, 30MeV; pulse repetitionrate, 30–100Hz; pulse width, 100 ns–3 ms; and average elec-tron current, 40–60 mA. The accelerated electrons struck awater-cooled Ta target, producing fast neutrons having theevaporation spectrum with an average energy of �0:7MeV.Slow neutrons were produced by moderating fast neutrons ina 10 cm thick water moderator coupled with a neutron-pro-ducing target. To decrease the overloading effect caused inthe detector phototubes by an intensive flash of gamma raysfrom the target, two kinds of target-moderator geometrieshave been tested. In the first case, a 10-cm-thick lead shadowbar was placed on the neutron beam axis close to the neu-tron-producing target in such a way that the sample couldnot see the target but only the moderator. In the second case,without the shadow bar, it was found that �30–40mm dis-
yOn leave from Petersburg Nuclear Physics Institute, Russia.
Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 42, No. 2, p. 135–144 (February 2005)
135
ORIGINAL PAPER
placement of the neutron-producing target from the neutronbeam axis effectively decreased the overloading effect of thegamma-flash while the neutron flux at the sample positionincreased significantly. Therefore, this second geometryhas been used for the measurements. The intensity of theneutron beam was monitored by a BF3 counter located ina gap between two sections of evacuated flight tubes.
A 24.2-m flight path equipped with evacuated flight tubeswas used. The neutron beam was collimated to 50mm diam-eter at the capture sample position by a set of lead, iron andH3BO3 collimators. The BGO detector was shielded againstexternal neutrons and gamma rays using borated paraffin andlead blocks of 5 or 10 cm thick. The distance between thecapture sample and the edge of the last section of the evac-uated flight tube was about 40 cm, and this part of the flightpath was not evacuated.
2. SamplesThe sample of 237Np was 1.13 g of neptunium oxide
(NpO2) powder packed in an aluminum disk container30mm in diameter and 0.4mm thick walls. The thicknessof the sample was equal to 0.359 g/cm2. The chemical purityof 237Np was 99.6% by weight, and the major impuritieswere Ga, K, P, Rb and S (�4mg of total weight). The radio-active purity (analyzed by alpha spectrometry by manufac-turer, Amersham) of the sample was 99.9% with an admix-ture of 238Pu less than 0.1%. The dummy sample—an iden-tical aluminum container without the neptunium oxide pow-der was used for the background measurement. This coupleof samples (237Np and dummy) was used in a previous meas-urement4) of capture cross sections carried out at the KURRILinac.
The incident neutron flux shape was measured with the10B sample made of boron powder encapsulated in a cylinderthin-walled (0.2mm) aluminum container 52mm in diame-ter, with a sample material thickness 0.494 g/cm2. The ele-mental purity of boron was >99:9%, while the 10B enrich-ment was 93%.
The capture samples were placed in the geometrical centerof the BGO detector using the suspenders—two aluminumstrips (3mm wide and 0.2mm thick) with a distance of32mm between them attached to the ring-shaped aluminum
holder. For the case of 237Np, a natural radioactivity(26MBq) of the sample material leads to the very hightime-of-flight independent background caused by 86.5 keVgamma-rays from 237Np, and 300, 312 and 341 keV gam-ma-rays from 233Pa (a daughter nucleus formed through �-decay of 237Np). To decrease this constant background, a7mm thick cylindrical lead shield was inserted in thethrough hole of the detector. It attenuated intensity of237Np gamma rays �8�106 times and those of 233Pa �14–30 times. The attenuation of 237Np capture gamma rays withaverage energy �1MeV by this shield was �1:5 only.
3. BGO Detector and Data Taking SystemThe capture gamma-ray detector shown in Fig. 2 consists
of 2 identical halves, each containing 8 optically separatedBGO crystals 170mm in length. These crystals form a cylin-drical scintillator layer having internal and external diame-ters of 110mm and 210mm, respectively. The total volumeof the detector scintillator is 8.54 l. Each BGO crystal isviewed by a 51mm � phototube.
A block diagram of the data taking system is shown inFig. 3. The outputs of 16 photomultiplier tubes (PMTs) werecombined into 2 groups and fed into two ORTEC 474 timingfilter amplifiers. After shaping and amplification, the signalswere summed and inverted by the Tennelec TC253 unit andfed into a 40MHz 4-channel 12-bit transient recorderPCI-412 (TRP-25). This flash-ADC board was installed ina 1.2 GHz personal computer (PC, Dell, model Dimension8200). A start signal for triggering the flash-ADC was gen-
Fig. 1 Scheme of the neutron capture cross section TOF-measurement at the 24.2m flight path at the KURRI Linac
Fig. 2 BGO detector: front view (left) and side view (right) with avertical cross section of one half. All sizes are given in mm
136 O. SHCHERBAKOV et al.
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erated by a linac electron pulse loaded on a Ta neutron-pro-ducing target.
The pulses from the BGO detector were digitized into25 ns time bins and temporarily stored in 2Mb buffer mem-ory of the flash-ADC. Then, after having been transferred tothe PC’s memory, each pulse was processed on-line to calcu-late its pulse height and time position using a simple and fastalgorithm. The raw data (input waveforms), as well as theTOF and pulse height spectra created on-line, were recordedon a hard disk of the PC. The subsequent analysis of the ac-cumulated raw data was carried out in off-line mode.
4. Measurement ProcedureThe measurements were performed using two different
modes of the Linac operation. For the measurement in theneutron energy interval 0.3 eV–1 keV, a short pulse width100 ns, repetition rate of 100Hz and average current of35 mA were used. For the measurement in the interval0.02 eV–1 keV, a long pulse width 3 ms, repetition rate50Hz and average current 70 mA were used. For both modesthe Linac was operated at the electron energy of 30MeV. Intotal, the measuring time was �50 h for the long pulse mode(about 32 h with the 237Np sample) and �110 h for the shortpulse mode (about 70 h with 237Np).
A highest possible 40MHz sampling frequency of flashADC corresponding to 25 ns TOF-channel width was usedfor all measurement runs to provide at least 10–12 channelsper pulse. As a result, the number of TOF channels used forthe short pulse mode to cover the neutron energy range downto 0.3 eV was 128 k, while for the long pulse mode (with thelowest neutron energy of 0.02 eV) it was equal to 512 k. Toprovide a meaningful statistic per channel, the TOF spectrawere re-binned to the final length of 8,192 channels, withthe channel width of 400 ns and 1.6 ms for short and longpulse mode, respectively. A 12-bit resolution (�2;048 chan-nels) of the flash ADC provided detailed information on thepulse height distribution of the detector pulses.
The measurements were carried out as a sequence of runsperformed with 237Np, dummy and 10B samples with andwithout black resonance filters used to evaluate the back-ground at various energy points. The ‘‘TOF-neutron energy’’calibration was made using the energies of resolved isolatedresonances of 237Np from ENDF/B-VI8) and JENDL-3.3.9)
The gamma ray pulse height scale was calibrated using
standard sources of 137Cs, 60Co and 480 keV gamma raysfrom neutron capture by 10B.
III. Data Processing and Analysis
1. Raw Data TreatmentA distinguishing feature of digital signal processing used
in present measurements, was a huge Gb-sized volume ofraw experimental data. As a result, the most time-consumingprocedure of the off-line data reprocessing was a calculationof the amplitude (pulse height) and time position of each de-tector pulse. It took a time about 2–3 times longer than a cor-responding measuring time because the more complicatedalgorithm of the pulse processing was used in the off-lineanalysis.
Each pulse was smoothed using the fast ‘‘zero-area’’ dig-ital filter.10) This algorithm produced a derived waveform,which has a zero value in the absence of peaks (signals)and has a shape similar to a smoothed second derivative inthe regions where the peaks are present. The wide (morethan few pulse widths) spectral structures were also removedfrom the waveform. This procedure was principally impor-tant because the base line of the detector output was shiftedby intense gamma-flash from the neutron-producing target.The results of this analysis were stored in a 2-dimensional‘‘TOF � pulse height’’, 8;192�2;048 matrix that was subse-quently used to obtain TOF and pulse height spectra for spe-cifically chosen amplitude or timing windows. By using con-ventional analog pulse processing, such a matrix of experi-mental data cannot be changed after an experiment has beencompleted. The digital pulse processing can be repeated anytime when it is necessary to use more sophisticated proce-dures to correct the data for pulse pile-up, base-line displace-ment and other distortion effects. Examples of the TOF spec-tra for 10B and 237Np samples obtained from the correspond-ing matrices by convolution along the pulse height axis with-in 300–800 keV and 1.3–5.8MeV windows, respectively, areshown in Figs. 4 and 5.
2. Backgrounds in TOF MeasurementsFor the 10B sample, the TOF (energy) dependence of the
background has been obtained by non-linear interpolationbetween the values observed in the transmission minimameasured with black resonance filters: 336 eV (Mn),132 eV (Co), 5.19 eV (Ag), 1.457 eV (In) and below�0:3 eV (Cd). It was found that using 4 isolated pointsand constant component below the Cd cut-off energy asinput data, the background can be described by a sum ofconstant and exponential functions
yBðtÞ ¼ a1 þ a2 expð�a3tÞ; ð1Þ
where a1, a2 and a3 are the fitting parameters, and t is theneutron TOF. It was supposed that TOF dependence of thebackground for the 10B sample was the same for both spectrameasured with and without resonance filters. A depression ofneutron flux caused by presence of the filters in neutronbeam was taken into account by correcting a fitting parame-ter a2 with a normalization coefficient which was calculatedas an average ratio of counting rate observed in a few be-
BGO( 8 sections )
ORTEC653
HV Supply
SumORTEC
474TF Amplifier
BGO( 8 sections )
ORTEC653
HV Supply
SumORTEC
474TF Amplifier
TENNELECTC 253
Sum.& Invert.
Attenuator Fast Discr.
Delay
FLASH ADCFAST Com Tec
TRP25
PCDELL
Dimension 8200
Start
Stop
e-
Linac
Ta target
Fig. 3 Block diagram of the data taking system
Measurement of Neutron Capture Cross Section of 237Np from 0.02 to 100 eV 137
VOL. 42, NO. 2, FEBRUARY 2005
tween-resonance points (10–100 eV) with and without thefilters. The percentage values of the evaluated backgroundat the black resonance positions and at the maximum ofthermal bump are shown in Fig. 4. It can be seen that thebackground equal to 3–4% at 100–400 eV increases up to10–16% with a decrease of neutron energy down to thermalrange.
For the 237Np sample, the measurement of the backgroundwas more complicated because of the low capture-to-scatter-ing cross section ratio 0.07–0.7 between resonances of 237Npand comparatively high sensitivity of the BGO detector toneutrons scattered in the sample.6) Under these circumstan-ces, besides the resonance filters, a few additional measure-ments have been done. The dummy sample was used tomodel the TOF dependence of the background with the de-tails, which cannot be revealed using standard notch-filtermethod with a limited number of black resonances. Themeasurements with a graphite sample and a sample holderhave been made to evaluate, respectively, the backgroundcomponents associated with neutrons scattered in the sampleand detector structure materials. To verify the backgroundfitting functions obtained as mentioned above, the numberof counts observed in the foreground measurements (237Npsample without filters) at 10–12 energy points correspondingto capture cross section minima of 237Np (1–3 barn) in theresolved resonance energy range 1–100 eV have been alsofitted. It was found that for the case of 237Np sample thebackground could be fitted by a sum of constant, powerand two exponential functions (9 fitting parameters ai intotal):
yNpðtÞ ¼ a1 þ a2t�a3 þ a4 expð�a5ðt � a6Þ2Þ
þ a7t0:5 expð�0:5ðlnðt=a8Þ=a9Þ2Þ:
ð2Þ
The third term (Gaussian-type function) describes the back-ground caused by neutrons with energies from 100 eV to1 keV scattered in the sample and captured in the BGO.6)
Its contribution can be seen in Fig. 5 as a bump-shapedstructure on a regular descending background fitting curvein 150–400 channel number range. The fourth term (‘‘log-normal’’-type function) describes the scattered neutronsbackground component at thermal energies below �0:3 eV.
3. Self-shielding and Multiple Scattering CorrectionsThe capture yield YNpðEÞ for the 237Np(n; �) reaction was
calculated from the 10B(n; �) reaction yield YBðEÞ, countingrates CNpðEÞ and CBðEÞ measured with 237Np and 10Bsamples and corrected for backgrounds, as follows:4)
YNpðEÞ ¼CNpðEÞCBðEÞ
YBðEÞ: ð3Þ
The 237Np capture cross section was calculated from themeasured yield using the relation:
��NpðEÞ ¼YNpðEÞNNptNp
FcNpðEÞ; ð4Þ
where NNp and tNp are the 237Np atomic density and samplethickness, ��NpðEÞ is the capture cross section, and FcNpðEÞ isthe correction function for neutron self-shielding and multi-ple scattering in the sample. The correction function FcNpðEÞ
TOF channel number
100 1000
Cou
nts
/ cha
nnel
10
100
1000
1000010B sample (foreground)10B sample with notch resonance filters of Mn, Co, Ag, In and CdBackground fit
336 eV (Mn)
132 eV (Co)
1.457 eV (In)
5.19 eV (Ag)
Thermal bump(Emax = 0.05 eV)
0.3 eV(Cd cut-off)
%3 %4
%10%16
%12
Fig. 4 The time-of-flight spectra of 10B in the energy range0:018 eV<En<750 eV:
10B sample without filters (dotted), 10Bsample with resonance filters of Mn, Co, Ag, In and Cd (solid),fitted background (dash-dotted)
The energies of ‘‘black’’ resonance minima and percentagevalues of the background are shown, TOF channel width is equalto 1.6 ms, Linac pulse width is equal to 3 ms.
Fig. 5 The time-of-flight spectra of 237Np: (upper) neptuniumsample (solid) and dummy sample (fit, dash-dotted), energyrange 0:3 eV<En<1:8 keV; (lower)
237Np sample (backgroundsubtracted), 4:9 eV<En<32 eV
The energies of the resonances are shown in eV, TOF channelwidth is equal to 400 ns, Linac pulse width is equal to 100 ns.
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JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
for the 237Np was calculated using the analytical method ofDresner.11) This method was improved by employment ofthe Pade approximation of special functions ð�; xÞ and�ð�; xÞ used for accurate calculation of the neutron cross sec-tions with accounting for Doppler broadening of neutron res-onances, as well as the interference between potential andresonance neutron scattering.12) With this approach, the Dop-pler-broadened cross sections of 237Np were reconstructed inresonance energy range below �100 eV using the evaluatedresonance parameters of ENDF/B-VI and JENDL-3.3 in-stead of the point-wise evaluated data. The calculated cor-rection function FcNpðEÞ was subsequently energy-broad-ened by the resolution function of the present TOF facility.6)
The latter was approximated by a Gaussian function that wascalculated by taking into account the moderator contributionand timing uncertainties: neutron burst width �n and channelwidth �ch (minimum 25 ns). The self-shielding and multiplescattering correction function FcNpðEÞ for the sample tem-perature 300K, neutron burst width 100 ns or 3 ms, and chan-nel width 400 ns or 1.6 ms (16 or 64 channels grouping, re-spectively) in the neutron energy range 0.01–100 eV, cor-rected for energy resolution, is shown in Fig. 6. As it canbe seen, the self-shielding effect is large for present sampleof 237Np, especially for the low energy resonances.
The reaction yield YBðEÞ for the 10B sample was calculat-ed by using Eq. (5):
YBðEÞ ¼ ð1� expð�NB�totB ðEÞtBÞÞ �
�n�B ðEÞ�totB ðEÞ
ðFcBðEÞÞ�1; ð5Þ
where NB and tB are the atomic density and thickness ofthe boron sample, �totB ðEÞ and �n�B ðEÞ are the total and
10B(n; �)7Li� reaction cross sections, respectively, andFcBðEÞ is the neutron multiple scattering correction. Theevaluated neutron cross sections of 10B and 11B (admixturenucleus) from JENDL-3.3 were used for the calculations.The correction function FcBðEÞ accounting for the multipleneutron scattering was calculated using the approximationproposed by Schmitt.13) The energy dependence of thereaction yield and the multiple-scattering correction isshown in Fig. 7.
4. Experimental UncertaintiesThe major contributions to the total uncertainties were due
to statistical errors, neutron self-shielding and multiplescattering corrections, the background subtraction, and theuncertainty of absolute normalization at thermal point.
The statistical uncertainties were calculated for the presentdata on a point-by-point basis. Because of the low countingstatistics observed between resonances, the data were com-bined in the groups of variable number of points to improvestatistical accuracy and simultaneously to preserve the reso-nance structure of the cross section. Figure 8 shows a per-centage of the capture cross section statistical uncertainties���=�� incorporated with the background subtraction uncer-tainties, separately for 10B and 237Np samples. These errors
Fig. 6 Neutron self-shielding and multiple scattering correctionfor the 237Np sample broadened by the TOF resolution function
Neutron Energy (eV)
0.01 0.1 1 10 100
Rea
ctio
n Y
ield
, Sca
tterin
g C
orre
ctio
n
0.80
0.85
0.90
0.95
1.00
1.05
1.10
Reaction YieldScattering Correction
Fig. 7 Reaction yield and multiple scattering correction functionfor the 10B sample
Neutron Energy (eV)
0.01 0.1 1 10 100
Rel
ativ
e U
ncer
tain
ty (
%)
1
10
100
10B sample237Np sample
Fig. 8 Relative partial ‘‘statistical+background subtraction’’ ex-perimental uncertainties for 10B and 237Np samples. Long(3 ms) Linac pulse measurements with 1.6 ms TOF channels, 50channel groups
Measurement of Neutron Capture Cross Section of 237Np from 0.02 to 100 eV 139
VOL. 42, NO. 2, FEBRUARY 2005
are given to the results of the measurements carried out withthe long pulse mode and combined in 50 TOF channelgroups.
The uncertainty of the self-shielding and multiple scatter-ing correction �FcNpðEÞ was evaluated by direct calculationof the FcNpðEÞ using an assumption that uncertainty of the237Np capture cross section is about 3.3%. This value isequal to the sum of the experimental error obtained by West-on and Todd3) for the thermal cross section and those for theareas of low energy resonances of 237Np. The energydependence of the �FcNpðEÞ calculated for TOF channelwidth 400 ns and energy-broadened with the resolution func-tion for Linac pulse width 100 ns, is similar to that of theFcNpðEÞ shown in Fig. 6 with a maximum magnitude of2.1% for the 0.49 eV resonance. The uncertainty of the�FcNpðEÞ due to the non-uniform thickness of the NpO2
sample was assumed to be negligible (1%). For the 10Bsample, an uncertainty of the calculated reaction yieldincluding the self-shielding and scattering correction wasevaluated by taking into account a deviation of the10B(n; �) cross section from the 1=v shape. This uncertaintywas negligible for neutron energies from 0.01 to 10 eV, andthen slowly increasing up to 0.25% at 100 eV.
The energy independent systematic uncertainty of abso-lute normalization was evaluated to be equal to 2.3%. It isthe variance of an interpolated value of 237Np capture crosssection at thermal point obtained by the least squares linearfit (in ‘‘log �–logE’’ scale) of the present experimental datain the energy range 0.02–0.1 eV. The error of the referencevalue of 181 b was not included in the total normalizationuncertainty.
IV. Results and Discussion
In the present experiment the shape of the 237Np capturecross section was measured as a function of neutron energy.To obtain absolute value of the capture cross section and tocompare the energy-dependent data with the previous resultsof other authors or evaluated libraries, the present data werenormalized to the reference value 181 b at 0.0253 eV fromthe ENDF/B-VI since other data have been also normalizedto this value. The results of the measurements from the short(100 ns) Linac pulse run were normalized between 0.3 and2 eV to the data obtained from the long (3 ms) Linac pulserun.
The normalization of the capture cross sections measuredby the TOF method at the thermal point is a commonly usedprocedure. However, there is a large discrepancy in the ther-mal capture cross sections of 237Np obtained in previousmeasurements. Therefore, the energy dependence of themeasured cross section is discussed first and thereafter aproblem of normalization.
The results of present measurements below 1 eV areshown in Fig. 9 together with the data by Weston andTodd,3) Kobayashi et al.,4) and the evaluated data ofJENDL-3.3 and ENDF/B-VI. Figures 10 and 11 show thepresent data and ENDF/B-VI in the energy ranges 0.02–100 eV and 15–58 eV. In the last two figures, the results ofpresent measurements were combined in 20 TOF channel
and 2 channel groups, respectively. For comparison, theevaluated data of ENDF/B-VI were also energy-broadenedby the resolution function of our spectrometer.
Neutron Energy (eV)
0.01 0.1 1
Cap
ture
Cro
ss S
ectio
n (b
arn)
10
100
1000
Weston & Todd (1981)Kobayashi et al. (2002)JENDL - 3.3ENDF/B - VI
Present (τn = 100 ns)
Present (τn = 3 µs)
Fig. 9 The 237Np capture cross section of present measurementsbelow 1 eV in comparison with the data by Weston and Todd,3)
Kobayashi et al.,4) and the evaluated data of JENDL-3.3 andENDF/B-VI
Neutron Energy (eV)
1 10 100
Cap
ture
Cro
ss S
ectio
n (b
arn)
1
10
100
1000
10000
ENDF/B - VI
Present (τn = 100 ns, τch = 400 ns x 20)
Fig. 10 The 237Np capture cross section of present measurementsfrom 0.2 to 100 eV in comparison with the evaluated data ofENDF/B-VI
Neutron Energy (eV)
20 30 40 50
Cap
ture
Cro
ss S
ectio
n (b
arn)
0
200
400
600
800
1000
1200
1400 ENDF/B - VI
Present (τn = 100 ns, τch = 400 ns x 2)
Fig. 11 The 237Np capture cross section of present measurementsfrom 15 to 58 eV in comparison with the evaluated data ofENDF/B-VI
140 O. SHCHERBAKOV et al.
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
For a quantitative comparison with the other data, the cap-ture cross sections from 0.02 to 100 eV measured in shortand long Linac pulse runs were averaged and then integratedwithin the energy intervals the same as those of Ref. 3).These average cross sections are given in Table 1 togetherwith the ratios of present data to analogous values by Westonand Todd, Kobayashi et al., and the evaluated data ofJENDL-3.3 and ENDF/B-VI. Figure 12 displays the aver-aged cross sections, and the ratios are shown in Fig. 13.
In the energy range below 0.1 eV, the present data agreewith the other data except JENDL-3.3 (due to normalizationof present results to the thermal cross section 181 b ofENDF/B-VI). In the energy interval 0.1–0.3 eV where aminimum in cross section exists, the best agreement is ob-served between the present data and ENDF/B-VI, whilethe data of Kobayashi et al. are �15% lower than present da-ta. In the range 0.3–0.85 eV containing the lowest-energyresonance 0.49 eV, the present data are �2:7% lower thanthose by Weston and both evaluations. The average crosssection by Kobayashi et al. is 17–20% above the present
and other data.For neutron energies between 1 eV and 100 eV, the pres-
ent average cross section is close to ENDF/B-VI andJENDL-3.3 except the energy intervals 2.8–5.2 eV and 28–45 eV where the present cross section is 4–8% above theevaluated data. In the energy range from 1 to 100 eV the en-ergy dependence of the present cross section is close to theexperimental data by Weston and Todd.3) However, thereis an obvious discrepancy between the data by Kobayashiet al.4) and present data.
In total, in the investigated energy range 0.02–100 eVthere is an agreement in the energy dependence of the237Np capture cross section between the present measure-ments and the ENDF/B-VI. As far as the JENDL-3.3 is con-cerned, there is an obvious disagreement with the present da-ta below �0:3 eV, while above this energy an agreement ex-ists, though it is worse than in a case of ENDF/B-VI. Itcould be stated that present experimental data do not supportthe conclusion made in Ref. 4) that there are somethingwrong with the values of the energy dependent capture cross
Table 1 Comparison of 237Np average capture cross sections obtained in present measurements with the results of otherauthors and evaluated data
Fig. 13 The ratios of the average capture cross section of 237Npfrom 0.02 to 100 eV obtained in present experiment to thosemeasured by other authors and evaluated data
Measurement of Neutron Capture Cross Section of 237Np from 0.02 to 100 eV 141
VOL. 42, NO. 2, FEBRUARY 2005
sections by Weston and Todd, and both of the evaluated databelow �0:2 eV.
It is also interesting to compare the partial resonance cap-ture integrals I0:3 (100 eV) and I0:5 (100 eV) calculatedusing the present and other experimental data, as well asevaluated capture cross sections of 237Np below 100 eV.The relevant data are given in Table 2. These data show thatpresent capture cross section weighted by a 1=E neutronspectrum from 0.3(0.5) to 100 eV is in overall agreementwith the data by Weston and Todd and both evaluations.
For comparison of resonance capture integrals obtainedfrom present data with the activation measurements, the par-tial resonance capture integral corresponding for energy in-terval 100 eV–10 keV was calculated using evaluated dataof ENDF/B-VI. This value equal to 62 barn was added ascorrection to the partial resonance integrals I0:3 (100 eV)and I0:5 (100 eV) given in Table 2. The results are shownin Table 3 in comparison with activation data by Schumanet al.,14) Hellstrand et al.,15) Eberle et al.,16) Kobayashiet al.,17) and Katoh et al.18) The present corrected valueobtained for resonance integral above 0.3 eV is in a goodagreement with the most recent value by Katoh et al.18)
For the lower boundary 0.5 eV, the present data agree wellwith those by Hellstrand et al.15) and Kobayashi et al.,17)
while there exists an obvious disagreement with Schumanet al.14) and Eberle et al.16)
For the case of the 237Np capture cross section measure-ment by the TOF method, besides the experimental difficultyof accurate evaluation of the background for a highly radio-active sample, there is also a problem of absolute normaliza-tion of the shape measurements. Among the previous meas-urements of the 237Np capture cross section, Kobayashiet al.4) normalized their data to the reference thermal valueof 181 b from the ENDF/B-VI. The authors of two othermeasurements2,3) normalized their data to the resonance pa-rameters of Mewissen et al.2) and Paya19) obtained from the
total cross section measurements in the energy range below100 eV. The problem of the normalization to thermal valuefor 237Np existed in the past because of a large discrepancy(158–187 b) between the thermal capture cross sections ob-tained mainly by activation method, as it was discussed inRefs. 3,4). The most recent value of the thermal capturecross section 141:7�5:4 b by Katoh et al.18) obtained alsoby the activation method even enlarges this discrepancy.
As far as the evaluated data are concerned, there is a largediscrepancy about 20 b between thermal capture crosssections 181 b and 161.7 b of ENDF/B-VI and JENDL-3.3,respectively. A recommended value 175:9�2:9 b by Mugh-abghab20) is between the evaluated ones. A value 181 b ofthe ENDF/B-VI is practically equal to 180�6 b by Westonand Todd3) which is based on the shape of their capture crosssection normalized by the resonance parameters. A value161.7 b of the JENDL-3.3 was calculated from the resonanceparameters obtained in the recent total cross section meas-urements by Gressier et al.21) This evaluated value is closeto recent activation data.
If the thermal value 141.7 b by Katoh et al.18) is used fornormalization instead of 181 b, then the present capture crosssection of 237Np is below the other experimental data andboth evaluations in the investigated energy range 0.02–100 eV. To perform such re-normalization, it is necessaryto calculate a new correction function for self-shieldingand multiple scattering. As a result, not only an absolutemagnitude of the capture cross section is changed but alsoa shape of the cross section is modified to some extent.The partial resonance integrals given in Table 2 after re-nor-malization decreased 1.342 times, while the reference ther-mal cross section decreased 1.277 times only. In Table 3, be-sides the resonance integrals calculated from the present datanormalized to 181 barn, also given are the data normalized to141.7 barn. A comparison of the present data normalizedboth to 181 and 141.7 b with the evaluated data for neutron
Table 2 Comparison of the 237Np partial capture resonance integral obtained in the present measurements with the otherTOF data and evaluations
Norm. (1): Normalized to the reference thermal cross section of 181 b
Norm. (2): Normalized to the reference thermal cross section of 141.7 b
142 O. SHCHERBAKOV et al.
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
energy from 0.02 to 1 eV is shown in Fig. 14. More than30% reduction of the capture cross section in resonancescaused by new normalization is in a strong contradictionwith the other data.
There is a possibility to solve a problem of the 237Np cap-ture cross section normalization at thermal point in a self-consistent manner by making use of the TOF method, with-out employment any other data for normalization. For thispurpose, it is necessary to perform precise total cross sectionmeasurements in thermal energy range and then to obtaincapture cross section as a difference of the total cross sectionand partial cross sections except the capture. Among thesecomponents, a scattering is the largest one, which could bemeasured separately or evaluated from the existing experi-mental data with a sufficient accuracy. For example, a differ-ence between evaluated scattering cross section values ofENDF/B-VI and JENDL-3.3 is less than 1 b in thermal en-ergy range 0.001–0.1 eV. The other contributions includingfission cross section are 1 b and could be neglected. Atpresent, only the total cross section data of Smith et al.22)
are available in the energy range 0.02–0.1 eV. An obviouscontradiction between these old experimental data and re-cent evaluations inspires a necessity of new measurementsof the 237Np total cross section below 1 eV.
V. Conclusion
The neutron capture cross section of 237Np has been meas-ured relative to the 10B(n; �)7Li� cross section by the time-of-flight method in the energy range from 0.02 to 100 eV.The electron linear accelerator at the Research Reactor Insti-tute, Kyoto University, was used as a pulsed neutron source.The 8.54-liter BGO scintillator was employed as a total ab-sorption detector for registration of prompt capture gammarays in conjunction with the flash ADC-based data takingsystem.
When the present data are normalized to the referencethermal value 181 b from ENDF/B-VI and averaged in com-paratively wide energy intervals, a good agreement with thedata of Weston and Todd and the evaluated data of ENDF/B-VI is observed in the whole investigated energy range, as
well as a satisfactory agreement with the JENDL-3.3 above�0:3 eV. If the most recent value of 141.7 b at 0.0253 eV ob-tained by activation method is used for normalization, thepresent data are in a contradiction with the other experimen-tal data3,4) and both evaluations. It means that a problem ofdiscrepancy between thermal capture cross sections of 237Npobtained by TOF method and activation still exists. It can besolved by making use of a high accuracy data to be obtainedfrom the total cross section TOF measurements below 1 eV.
Acknowledgments
The authors would like to express their gratitude to theLinac staff of the Research Reactor Institute, Kyoto Univer-sity for cooperation in the measurements and steadily oper-ation of the Linac. Help and participation of K. Takami andSamyol Lee in present research are highly appreciated.
The present investigation was carried in part within theframework of the JNC International Fellowship ResearchProgram.
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Fig. 14 Comparison of present data normalized to 181 and141.7 b at thermal energy 0.0253 eV with the evaluated data ofJENDL-3.3 and ENDF/B-VI
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