Measurements of the ²⁴³Am neutron capture and total cross sections with ANNRI at J-PARC

ABSTRACT

Neutron total and capture cross sections of ²⁴³Am have been measured in Accurate Neutron Nucleus Reaction measurement Instrument at Materials and Life Science Experimental Facility of Japan Proton Accelerator Research Complex with a neutron time-of-flight (TOF) method. The neutron capture cross section in the energy region from 10 meV to 100 eV was determined using an array of Ge detectors. Three samples with different activities were used for measurements of the capture cross section. The neutron total cross section in the energy region from 4 meV to 100 eV was measured using Li-glass detectors. Derived cross-section value at neutron energy of 0.0253 eV is $87.7\pm 5.4$ b for the capture cross section and $101\pm 11$ b for the total cross section.

KEYWORDS:

• ²⁴³am
• neutron capture cross section
• neutron total cross section
• J-PARC
• MLF
• ANNRI
• neutron TOF

1. Introduction

Nuclear data for minor actinides (MAs) and long-lived fission products are necessary for burn-up analyses of nuclear fuels, improving the design and performance of advanced nuclear reactors and the nuclear transmutation [1–4]. In particular, ²⁴³Am is one of the important MAs in the study of nuclear transmutation because ²⁴³Am contributes most to the total radiotoxicity of the spent fuel for thousands of years because of its long half-life (7370 yr). The cross section of ²⁴³Am(n,$\gamma$) is also important because it produces ²⁴⁴Cm, which is a strong neutron emitter, and also whose yield determines the productions of higher-mass Cm isotopes, e.g. ²⁴⁵Cm (8500 yr) and ²⁴⁶Cm (4760 yr).

The capture and transmission differential data available in the literature for ²⁴³Am are summarized in Table 1. The latest data are given by Mendoza et al. in 2014 measured at the neutron time-of-flight (n_TOF) facility at CERN [5]. The result covered the neutron energy region from 0.7 eV to 2.5 keV and did not cover the thermal energy region. There are one capture cross section measurement by Hori et al. [6] and two transmission measurements by Berreth et al. [7] and Cote et al. [8] covering the thermal energy region. However, these two transmission measurements were presented in 1970s. The experimental data available for the evaluation of the ²⁴³Am capture cross-section at thermal energies are summarized in Table 2. As it can be observed, the data are scattered mainly between 75 and 85 barns, and are in some cases incompatible. There is no clear choice about which thermal capture cross section should be used from the point of view of the data.

Table 1. Differential capture and transmission measurements of ²⁴³Am

Reference	Year	Type	Energy range
Mendoza et al. [5]	2016	Capture	0.7 eV–2.5 keV
Hori et al. [6]	2009	Capture	10 meV–400 eV
Jandel et al. [9]	2009	Capture	8 eV–250 keV
Weston et al. [10]	1985	Capture	258 eV–92.1 keV
Wisshak et al. [11]	1983	Capture	5 keV–250 keV
Belanova et al. [12]	1976	Transmission	0.35 eV–35 eV
Simpson et al. [13]	1974	Transmission	0.5 eV–1 keV
Berreth et al. [7]	1970	Transmission	8 meV–25.6 eV
Cote et al. [8]	1959	Transmission	1.4 meV–15.44 eV
This Work		Capture	10 meV–100 eV
		Transmission	4 meV–100 eV

Table 2. Thermal capture cross sections ${\sigma }_0$ provided by different integral measurements and evaluations of ²⁴³Am

Reference	Year	${\sigma }_0$(b)
This Work	2019	$87.7\pm 5.4$
Hori et al. [6]	2009	76.6
Marie et al. [14]	2006	$81.8\pm 3.6$
Hatsukawa et al. [15]	1997	84.4
Gavrilov et al. [16]	1977	$83\pm 6$
Eberle et al. [17]	1971	$77\pm 2$
Folger et al. [18]	1968	78
Bak et al. [19]	1967	$73\pm 6$
Ice et al. [20]	1966	66–84
Butler et al. [21]	1957	$73.6\pm 1.8$
Harvey et al. [22]	1954	$140\pm 50$
Stevens et al. [23]	1954	115
ENDF/B-VII.1 [24]	2011	80.4
JENDL-4.0 [25]	2011	79.3

To improve the situation, capture cross section of ²⁴³Am was measured in Accurate Neutron Nucleus Reaction measurement Instrument (ANNRI) at Materials and Life Science Experimental Facility (MLF) of Japan Proton Accelerator Research Complex (J-PARC) with a neutron time-of-flight (TOF) method. Because capture cross section of ²⁴³Am is more than 90% of total cross section at thermal energy, total cross section of ²⁴³Am was also measured to check consistency of measured cross sections. Because uncertainty due to the sample mass is one of the most important factors contributing to the discrepancies of measured cross section data [26,27], the activities of the samples were determined accurately with a calorimetric method and a gamma-ray spectroscopic method [28].

The outline of the neutron TOF experiment facility, ANNRI, and information of the samples are described in Section 2. The measurement of capture cross section is presented in Section 3, followed by the neutron total cross section measurement in Section 4. At the end, conclusion is given in Section 5.

2. Facilities and samples

2.1. ANNRI at J-PARC

The experiments with the neutron TOF method were performed in ANNRI. Pulsed neutrons were produced in spallation reactions in the mercury target of the MLF by 3-GeV pulsed protons from the 3-GeV rapid cycling synchrotron in J-PARC [29]. Produced neutrons were slowed down in a supercritical hydrogen moderator and transported to ANNRI, which is designed for neutron TOF measurements.

Layout of ANNRI is shown in Figure 1. Two gamma-ray spectrometers and one neutron detector system are installed in ANNRI. In this measurement, an array of large Ge detectors installed at flight length of 21.5 m and Li-glass detectors installed at flight length of 28.7 m were used. In ANNRI, neutron collimators, resonance filters, and chopper systems are installed to deliver an intense neutron beam to detector systems with good quality [30]. The neutron chopper systems were used to reduce influence of slow neutrons which were produced in the previous proton pulse and overlapped with neutrons produced in the latest pulse (frame-overlap neutrons). More details about the beam line are given in Ref [37].

Figure 1. Vertical cross sectional view of ANNRI.

In the experiments, J-PARC was operated under 25-Hz repetition and under a proton beam power of 200 kW. The proton intensity per shot was stable within 1% in FWHM. Each incident proton pulse consisted of $1.79\times {10}^{10}$ protons.

2.2. Samples

For the experiments, three ²⁴³Am samples with nominal activities of 60, 120, and 240 MBq were prepared. The samples were manufactured by V.G. Khlopin Radium Institute in Russia. Americium dioxide (AmO₂) powder was mixed with Y₂O₃ powder and pressed into pellets with diameter of 10.0 $\pm$ 0.1 mm and a thickness of 0.5 $\pm$ 0.1 mm, according to the specifications provided by the manufacturers. The pellets were encapsulated in Al containers with a diameter of 22 mm and a thickness of 0.1 mm. Similarly, a ‘dummy sample’ which only contains Y₂O₃ powder without AmO₂ powder was prepared. The weight of Y₂O₃ powder of each sample is shown in Table 3.

Table 3. Characteristics of the samples

	Activities measured (MBq)
Samples	Calorimetric method	$\gamma$-ray spectrometry	Weight of Y2O3 [mg]
60-MBq ^243Am Sample	$67.3\pm 0.3$	$66.7\pm 1.3$	45
120-MBq ^243Am Sample	$155.8\pm 0.3$	$155\pm 3$	43
240-MBq ^243Am Sample	$281.8\pm 0.3$	$286\pm 6$	39
Dummy Sample	-	-	66

The activities of the Am samples were determined with a calorimetric method and gamma-ray spectroscopic method independently [28]. In the calorimetric method, decay heat of the ²⁴³Am samples was measured accurately by using the calorimeter with a precision of 200nW, and thus activities of the samples were obtained accurately with uncertainties less than 0.45%. For the gamma-ray spectroscopic method, accurate gamma-ray emission probabilities of ²⁴³Am and ²³⁹Np were determined by combination of gamma- and alpha-ray spectroscopy with uncertainties less than 1.2% and the obtained gamma-ray emission probability of 277.60 keV from ²³⁹Np (14.34 ± 0.16%) was used for activity determination [31]. The determined activities of the ²⁴³Am samples were obtained with uncertainties less than 2%. The activities of the ²⁴³Am samples obtained by the two different methods are shown in Table 3. Their results were in good agreement. Because the results by the calorimetric method have smaller uncertainties, they were used in this paper.

To make detailed isotopic and impurity analyses, liquid sample from the same batch of the samples was also supplied by the manufacturers. Isotopic and impurity abundances were determined by thermal ionization mass spectrometry (TIMS) and alpha-ray spectrometry [32]. The results are described in Table 4. The determined isotopic purity of ²⁴³Am was 97.67% with contamination of 2.29% and 0.04% by ²⁴¹Am and ²⁴⁴Cm, respectively. In this paper, the results with TIMS were used for Am isotopes, and those by alpha-ray spectrometry were used for ²⁴⁴Cm.

Table 4. Isotopic and impurity abundances of ²⁴³Am sample

	TIMS (%)	α-ray spectrometry (%)
^241Am	$2.29\pm 0.02$	$2.4\pm 0.4$
^242mAm	$0.038\pm 0.003$	Not Detected
^243Am	$97.67\pm 0.02$	$97.6\pm 4.0$
^244Cm	Not detected	$0.036\pm 0.001$

3. Neutron capture cross section measurements

3.1. Experimental procedure

For the neutron capture cross-section measurements, the array of Ge detectors was used to measure emitted gamma-rays from neutron capture reactions. The array of Ge detectors is composed of two cluster-Ge detectors, eight coaxial-Ge detectors and anti-coincidence shields around each Ge detector as shown in Ref [33]. Each cluster Ge detector consists of seven Ge crystals. The positions of the cluster Ge detectors were arranged so that the distance between the front surface of each detector and the center of the sample position was 175 mm. Lead shield with a thickness of 5 cm was placed in front of each cluster Ge detector to attenuate intense decay gamma-rays form ²⁴³Am and ²³⁹Np, which is daughter nuclide of ²⁴³Am. In the measurements, two cluster Ge detectors were used, but the coaxial Ge detectors were not used because they suffered from severe electrical noise. The signals from the Ge detectors were analyzed using CAEN V1724 (14bit 100 MHz) module with a digital pulse processing software, DPP-PHA [34]. A pulse height and a TOF were recorded in ‘event-by-event’ mode with the module.

The neutron beam was collimated to diameters of 22 mm at the sample position. To reduce the scattering of neutrons by air, helium gas ran through the beam duct.

In the capture cross-section measurement, the 60-, 120-, and 240-MBq ²⁴³Am samples were used. Each sample was put in a bag made of fluorinated ethylene propylene (FEP) films and attached to a sample holder. The sample holders were made of polytetrafluoroethylene and shaped into a 50 mm × 70 mm rectangular frame, which was made much larger than the size of the collimated neutrons with a diameter of 22 mm [30]. To estimate influence due to scattered neutrons, a carbon sample was used in Ref [26]. However, an enriched ²⁰⁸Pb sample with a diameter of 5.0 ± 0.1 mm and a weight of 170.2 ± 0.1 mg was used in this experiment because ²⁰⁸Pb and ²⁴³Am has similar mass number and capture cross section of ²⁰⁸Pb is much lower than that of carbon. For the background estimation, measurements with the dummy sample and the FEP bag without any sample (Blank) were also carried out. To obtain the energy dependence of the incident neutron flux, measurement with an enriched ¹⁰B sample with a diameter of $5.0\pm 0.1$ mm and a weight of $1.93\pm 0.01$ mg was also carried out.

The measuring times are summarized in Table 5.

Table 5. Measuring times

Samples	Measuring time [h]
60-MBq ^243Am Sample	9
120-MBq ^243Am Sample	7.5
240-MBq ^243Am Sample	13
Dummy sample	8
Blank	8
^208Pb	12
^10B	3

3.2. Data analysis

3.2.1. Pulse height spectra

Two prompt gamma-ray pulse-height spectra with the 240-MBq ²⁴³Am sample and the dummy case are shown in Figure 2. The spectra were measured with one of the 14 Ge crystals. They were normalized with the number of incident proton pulses. A decay gamma-ray pulse-height spectrum per repetition period of 40 ms (without neutron beam) is also shown in the figure. A difference between the ²⁴³Am and dummy spectra appears clearly below the Q-value of ²⁴³Am(n,$\gamma$) reaction (5363 keV). However, almost all the intense photo peaks shown in the spectrum of the 204-MBq ²⁴³Am sample are also observed in the spectra of dummy case and the decay gamma-ray. Prompt gamma rays from ²⁷Al (n,$\gamma$)²⁸Al, ⁵⁶Fe (n,$\gamma$) ⁵⁷Fe and ⁷³Ge (n,$\gamma$) ⁷⁴Ge reactions, and decay gamma rays from ²⁸Al are observed in both the gamma-ray spectra of the ²⁴³Am sample and the dummy sample. Peaks originating from ²⁴³Am neutron-capture events were hidden under the intense background gamma rays.

Figure 2. Prompt gamma-ray pulse-height spectra of the 240-MBq ²⁴³Am sample, the dummy case, and a decay gamma-ray pulse-height spectrum from the 240-MBq ²⁴³Am sample without neutron beam measured with one of the cluster crystals. Each spectrum was normalized with the number of incident proton beam shots.

3.2.2. Raw TOF spectra

As shown in Figure 2, although the lead shield with a thickness of 5 cm was placed, intense decay gamma-rays form ²⁴³Am and ²³⁹Np were observed, especially under 200 keV. In the energy range over 6000 keV, which is higher than the neutron separation energies of ²⁴⁴Am (5363.7 keV [35]), many intense photo peaks due to background gamma-rays are observed. In the analysis, to improve the signal-to-noise ratio, one gate from 600 keV to 6000 keV was applied to derive TOF spectra. Gated TOF count rate spectra per $1.79\times {10}^{10}$ protons of the 240-, 120-, and 60-MBq ²⁴³Am samples and the dummy sample are shown in Figure 3. The TOF spectra were measured with one of the 14 crystals. In the spectrum of the ²⁴³Am samples, not only resonances of the ²⁴³Am but also the resonances of ²⁴¹Am (2.82 ms = 0.305 eV, 2.06 ms = 0.572 eV) and ²⁴⁰Pu (1.519 ms = 1.056 eV) are also clearly observed. Therefore, to obtain neutron capture cross sections of ²⁴³Am, the contributions from these impurities must be evaluated and subtracted. The broad peak around 20 ms is due to the neutron bump thermalized in the liquid hydrogen in the moderator.

Figure 3. Gated TOF spectra per $1.79\times {10}^{10}$ protons of the ²⁴³Am samples and the dummy sample. The spectra were measured with one of the 14 crystals.

3.2.3. Dead time correction

Dead-time correction was applied to the gated TOF spectra with extended dead-time model using a dead time of 5.67 ± 0.02 $\mu$s (567 ± 2 ch.) per event due to the settings of the CAEN V1724 module. Figure 4 shows a typical time dependence of the calculated dead-time correction factor for one of the Ge crystals. The dead time correction factors at strongest resonance peak of ²⁴³Am ($1.34\mathrm{m}\mathrm{s}=1.356\mathrm{e}\mathrm{V}$) were 4.0% for the 240-MBq ²⁴³Am sample, 3.2% for the 120-MBq sample, and 1.8% for the 60-MBq sample. Those at neutron energy of 0.0253 eV were less than 0.4%. The uncertainties of the correction were calculated using the uncertainty of the dead time per event. This dead-time correction was done for each crystal because event rates depend on the positions of the crystals.

Figure 4. Typical time dependences of the dead-time correction factor for the measurements of the 240-MBq ²⁴³Am sample and the 60-MBq ²⁴³Am sample.

3.2.4. Background subtraction

In this analysis, we separately estimated backgrounds caused by frame overlapping neutrons and decay γ-rays, sample independent background, backgrounds due to aluminum container and Y₂O₃ included in the ²⁴³Am samples, and the backgrounds due to scattered neutrons by ²⁴³Am.

The backgrounds caused by frame overlap neutrons and decay gamma-rays were estimated by fitting the TOF spectra in the TOF region from 37 to 40 ms by using the following equation (an exponential decay plus constant function):

(1) $f\left(t\right)=a+b\times exp\left(-\frac{t-40ms}{t_c}\right)$(1)

where t is TOF and a, b, and $t_c$ are fitting parameters. The fitted function was moved back in time by 40 ms. This fitting was done for each sample and crystal. In this estimation, propagations of fitting uncertainties were taken into account.

The sample-independent background is represented by the TOF spectrum of blank. After subtracting the backgrounds caused by overlapping neutrons and decay γ-rays, the sample-independent background was subtracted from each spectrum.

The backgrounds due to aluminum container and Y₂O₃ included in the ²⁴³Am samples were derived from the spectrum of the dummy sample considering differences in Y₂O₃ masses between the ²⁴³Am samples and the dummy samples.

There still remains the background due to scattered neutrons by ²⁴³Am. This background was estimated by the spectrum of the ²⁰⁸Pb sample by considering the differences in the elastic-scattering cross section and the area densities of the samples. The subtraction procedure was the same manner as that described in Ref [26]. The elastic-scattering cross section was taken from the JENDL-4.0. The difference of diameters between the ²⁴³Am samples and ²⁰⁸Pb sample was not considered, because the diameter of the neutron beam at the sample position, 22 mm, was enough large in comparison to the diameters of the samples.

The estimations and subtractions of the backgrounds were done separately for each crystal, because the backgrounds depend on the positions of the crystals. On the other hand, procedures in the following steps are independent of the positions. Therefore, from here on, all 14 TOF spectra for each sample were summed up. The deduced net spectra of the ²⁴³Am samples are shown in Figure 5.

Figure 5. Deduced net spectra of the 240-MBq ²⁴³Am sample and that of the 60-MBq ²⁴³Am sample.

The TOF parameters for the neutron energy calibration were derived from a TOF spectrum of a gold sample and the resonance energies of gold listed in JENDL-4.0.

3.2.5. Correction for the neutron self-shielding and multiple-scattering effects

Corrections for the neutron self-shielding and multiple-scattering effects in the ²⁴³Am samples were made by using the PHITS code [36]. The shape, dimension, isotope abundances, and weight of the samples were taken into account in the calculation. Figure 6 shows the determined correction factors of the ²⁴³Am samples. The calculated correction factors at the strongest resonance were 1.29, 1.74, and 2.49 for the 60-, 120-, and 240-MBq ²⁴³Am samples, respectively.

Figure 6. Correction factors for neutron self-shielding and multiple scattering on the 240-MBq ²⁴³Am sample and the 60-MBq ²⁴³Am sample.

The calculated correction factors strongly depend on the areal densities of the ²⁴³Am. The areal densities were calculated from the activities of the samples (67.3 ± 0.3 MBq, 155.8 ± 0.3 MBq, 281.8 ± 0.3 MBq), half-life of ²⁴³Am (7357 ± 23y) [38], and diameter of the sample pellet (10.0 ± 0.1 mm). The uncertainties of the areal densities mainly come from the uncertainty of the pellet diameter. Therefore, uncertainties of the correction factors were derived from the uncertainty of the pellet diameter. To derive the uncertainties, correction factors for virtual samples were calculated. The virtual samples have same activities as the real samples and have pellets with diameter of 9.9 mm. Differences between the correction factors for the real samples and those for the virtual samples were considered as the uncertainties due to the correction.

3.2.6. Calculation of relative yield

In the present analysis, the incident neutron energy spectrum was determined by measuring 478-keV γ-rays emitted in ¹⁰B(n,αγ) reactions. TOF spectra gated on the 478-keV photo peak area were made for the ¹⁰B, ²⁰⁸Pb, and blank measurements. Backgrounds were estimated and subtracted in the same manner as the ²⁴³Am samples. The ¹⁰B(n,αγ) cross section listed in JENDL-4.0 was used to convert the detected counts to the number of the incident neutrons. The correction of the neutron self-shielding and multiple scattering was applied. The uncertainty of the corrections was deduced from the uncertainties of the ¹⁰B sample diameter. The obtained relative neutron flux and uncertainties are shown in Figure 7.

Figure 7. Relative neutron spectrum obtained using the ¹⁰B(n,$\alpha \gamma$)⁷Li reactions and uncertainties (total, statistical uncertainty, uncertainty due to self-shielding and multiple-scattering correction).

The relative capture yields of the ²⁴³Am samples were calculated from the deduced net spectra of the ²⁴³Am samples shown in Figure 5, the neutron self-shielding and multiple-scattering correction factors shown in Figure 6, and the averaged neutron flux shown in Figure 7.

3.2.7. Absolute capture cross section at 1.356 eV resonance

Absolute capture cross section at the 1.356 eV (= 1.34 ms) resonance was derived by taking ratios of the deduced capture yield before correction for the neutron self-shielding and multiple-scattering effects as shown in Figure 5.

In the measurements, capture yield of ²⁴³Am, $Y\left(E_n\right)$, is described as follows:

(2) $Y\left(E_n\right)=A\left(E_n\right)\times \phi \left(E_n\right)\times k_{Am}\times \frac{{\sigma }_{cap}\left(E_n\right)}{\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)+{\sigma }_{fission}\left(E_n\right)\right)\times N_{Am}+{\sum }_x{\sigma }_x\left(E_n\right)\times N_x}$(2)

($x=$ ²⁴¹Am, ²⁴⁴Cm, Al, Y, O),

where ${ }_n$ is neutron energy, $A\left(E_n\right)$ is correction factor due to multiple scattering, $\phi \left(E_n\right)$ is neutron flux, $k_{Am}$ is an averaged detection efficiency of the array of Ge detectors for neutron capture reactions of ²⁴³Am, ${\sigma }_{cap}\left(E_n\right),{\sigma }_{el}\left(E_n\right)$ and ${\sigma }_{fission}\left(E_n\right)$ are capture, elastic scattering and fission cross section of ²⁴³Am, $N_{Am}$ is areal density of ²⁴³Am, ${\sigma }_x\left(E_n\right)$ is total cross section of nuclide x, and $N_x$ is areal density of nuclide x. Because capture cross section is by 1000 times larger than fission cross section, influence due to fission events was negligible. In this calculation, $\sum _x{\sigma }_x\left(E_n\right)\times N_x$ was also negligible, because the products of the total cross sections and the areal densities for ²⁴¹Am, ²⁴⁴Cm, Al, Y, and O are much smaller than that for ²⁴³Am. Therefore, at the resonance, Equation (2)(2) $Y\left(E_n\right)=A\left(E_n\right)\times \phi \left(E_n\right)\times k_{Am}\times \frac{{\sigma }_{cap}\left(E_n\right)}{\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)+{\sigma }_{fission}\left(E_n\right)\right)\times N_{Am}+{\sum }_x{\sigma }_x\left(E_n\right)\times N_x}$(2) is described as follows:

(3) $Y\left(E_n\right)\cong A\left(E_n\right)\times \phi \left(E_n\right)\times k_{Am}\times \frac{{\sigma }_{cap}\left(E_n\right)}{{\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)}\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am}\right)\right).$(3)

Using Equation (3)(3) $Y\left(E_n\right)\cong A\left(E_n\right)\times \phi \left(E_n\right)\times k_{Am}\times \frac{{\sigma }_{cap}\left(E_n\right)}{{\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)}\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am}\right)\right).$(3) , the ratios of the capture yields of the 60-MBq and 120-MBq samples to that of the 240-MBq sample, $R_{60/240}\left(E_n\right)$ and $R_{120/240}\left(E_n\right)$, are described as:

(4) $\begin{array}{rl}& R_{60/240}\left(E_n\right)=\frac{Y_{60}\left(E_n\right)}{Y_{240}\left(E_n\right)}=\frac{A_{60}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}60}\right)\right)}{A_{240}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}240}\right)\right)}\\ & R_{120/240}\left(E_n\right)=\frac{Y_{120}\left(E_n\right)}{Y_{240}\left(E_n\right)}=\frac{A_{120}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}120}\right)\right)}{A_{240}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}240}\right)\right)}\end{array}$(4)

As shown in Equations (4(4) $\begin{array}{rl}& R_{60/240}\left(E_n\right)=\frac{Y_{60}\left(E_n\right)}{Y_{240}\left(E_n\right)}=\frac{A_{60}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}60}\right)\right)}{A_{240}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}240}\right)\right)}\\ & R_{120/240}\left(E_n\right)=\frac{Y_{120}\left(E_n\right)}{Y_{240}\left(E_n\right)}=\frac{A_{120}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}120}\right)\right)}{A_{240}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}240}\right)\right)}\end{array}$(4) ), by taking ratios of capture yields of the samples, neutron intensity and detection efficiency of the spectrometer are cancelled out.

The correction factors due to multiple scattering, $A_{60}\left(E_n\right)$, $A_{120}\left(E_n\right)$, and $A_{240}\left(E_n\right)$ were calculated with PHITS. The calculated values at the top of the resonance were 1.006, 1.007, and 1.005 for the 60-, 120- and 240-MBq ²⁴³Am samples, respectively. The areal densities ($N_{Am\mathrm{\_}60}$, $N_{Am\mathrm{\_}120}$, and $N_{Am\mathrm{\_}240}$) were obtained from the activities of the samples, half-life of ²⁴³Am (7357 ± 23y) [38], and diameter of the samples (10.0 ± 0.1 mm). The ratio of the capture yields between the samples, $R_{60/240}\left(E_n\right)$, and $R_{120/240}\left(E_n\right)$ is obtained from the measurements as shown in Figure 8. Therefore, in Equations (4(4) $\begin{array}{rl}& R_{60/240}\left(E_n\right)=\frac{Y_{60}\left(E_n\right)}{Y_{240}\left(E_n\right)}=\frac{A_{60}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}60}\right)\right)}{A_{240}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}240}\right)\right)}\\ & R_{120/240}\left(E_n\right)=\frac{Y_{120}\left(E_n\right)}{Y_{240}\left(E_n\right)}=\frac{A_{120}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}120}\right)\right)}{A_{240}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}240}\right)\right)}\end{array}$(4) ), all the parameters except for ${\sigma }_{cap}\left(E_n\right)$ and ${\sigma }_{el}\left(E_n\right)$ are known. ${\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)$ is obtained from Equation (4)(4) $\begin{array}{rl}& R_{60/240}\left(E_n\right)=\frac{Y_{60}\left(E_n\right)}{Y_{240}\left(E_n\right)}=\frac{A_{60}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}60}\right)\right)}{A_{240}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}240}\right)\right)}\\ & R_{120/240}\left(E_n\right)=\frac{Y_{120}\left(E_n\right)}{Y_{240}\left(E_n\right)}=\frac{A_{120}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}120}\right)\right)}{A_{240}\left(E_n\right)\times \left(1-\mathrm{e}\mathrm{x}\mathrm{p}\left(-\left({\sigma }_{cap}\left(E_n\right)+{\sigma }_{el}\left(E_n\right)\right)\times N_{Am\mathrm{\_}240}\right)\right)}\end{array}$(4) . Absolute neutron capture cross sections of ²⁴³Am around the 1.356 eV resonance, ${\sigma }_{cap}\left(E_n\right)$, were obtained by subtracting the elastic scattering cross section data in JENDL-4.0 (${\sigma }_{el}\left(E_n\right)$).

Figure 8. Ratios of the capture yields of the 120-MBq ²⁴³Am sample and the 60-MBq ²⁴³Am sample to that of the 240-MBq ²⁴³Am sample.

Figure 9 shows the obtained capture cross sections together with values of JENDL-4.0 at 300 K (broadening with the resolution function of ANNRI [39]). The obtained cross sections are consistent with the evaluated data in JENDL-4.0.

Figure 9. Determined neutron capture cross sections of ²⁴³Am around the 1.356 eV resonance together with values of JENDL-4.0 for temperature of 300 K (broadened with the resolution function).

In this analysis, the statistical uncertainty, the uncertainty due to area densities, and the uncertainty of the evaluated elastic scattering cross section were taken into consideration. The derived uncertainties are shown in Figure 10. Because capture cross section is about 50 times larger than the elastic scattering cross section, influence due to the uncertainty of the evaluated elastic scattering cross section was much smaller than the other uncertainties. The main component of the uncertainty comes from the uncertainties of the areal densities. The uncertainties due to the areal densities of the ²⁴³Am samples were derived from the uncertainties of the sample activity, half-life of ²⁴³Am, and the diameter of the sample. The main component of the uncertainties due to the areal densities comes from the uncertainty of the sample diameter.

3.2.8. Evaluation of influence of fission events

At the 1.356-eV resonance of ²⁴³Am, capture cross sections of ²⁴³Am are much larger than fission cross sections of impurities in the sample, and influence of fission event is negligible. However, in the neutron energy region below 1 eV, the influences of fission events had to be evaluated and subtracted, because the samples contain ²³⁹Pu (see Sec. 3.2.10) and ²³⁹Pu has large fission cross section in this region. The ratio of the sensitivity of fission events to capture events for the array of Ge detectors had been deduced in Ref [33]. The deduced ratio was 2.3 ± 1.0. This means fission events are 2.3 times more sensitive than capture events for the spectrometer. In this paper, the same value was used to estimate the influence of fission events.

3.2.9. Normalization

The relative yield of the 60-MBq ²⁴³Am sample calculated in Sec. 3.2.5. was normalized to the derived capture cross section shown in Figure 10(a) using the area of the 1.356 eV resonance. Contaminations of the fission events are small and negligible, because the neutron-capture cross section of ²⁴³Am at the resonance is about 2000 times larger than the fission cross section.

The relative yields of the 120-MBq and 240-MBq ²⁴³Am samples were normalized at the 1.744-eV resonance to the normalized yield of the 60-MBq ²⁴³Am sample because the influences of the self-shielding and multiple-scattering correction on the 120-MBq and 240-MBq ²⁴³Am samples for the 1.356 eV resonance were too large as shown in Figure 6. The normalized yields from the measurements of the ²⁴³Am samples around 1.356 and 1.744-eV resonances are shown in Figure 11. The three normalized yields agreed well.

Figure 11. Normalized yield from the measurements of the ²⁴³Am samples around 1.356- and 1.744-eV resonances.

Uncertainties due to the normalization were derived from the uncertainty of the derived absolute cross section shown in Figure 10, the uncertainties due to the self-shielding and multiple-scattering correction, and the uncertainties due to the area of the resonances. The deduced uncertainties were 3.3% for the 60-MBq sample measurement and 3.9% for the 120-MBq and 240-MBq sample measurements.

Figure 10. Derived cross section, total uncertainty, statistical uncertainty, uncertainty due to area densities, and uncertainty due to values of evaluated elastic scattering cross section.

3.2.10. Evaluation of the abundance of Pu isotopes

Figure 12 shows the normalized yields from the measurements of the 60-MBq and 240-MBq samples, and comparison to contribution from ²⁴³Am and known impurities which was calculated using JENDL-4.0 and abundance ratios written in Table 4.

Figure 12. The normalized yields from the measurements of the 240-MBq samples; comparison to the contribution from ²⁴³Am and Cm impurities and contribution from deduced impurities of ²⁴⁰Pu and ²³⁹Pu. The contributions were calculated using JENDL-4.0.

Although, the isotope ²⁴⁰Pu is not included in the isotope and impurity analyses given in Table 4, the 1.056-eV resonance of ²⁴⁰Pu is clearly observed. The abundance of ²⁴⁰Pu was derived using an area of the 1.056-eV resonance for the measurement of the 240-MBq sample and evaluated cross section in JENDL-4.0. The derived abundance of ²⁴⁰Pu was 0.77 ± 0.04%.

The resonance around 0.30 eV is observed in Figure 12. There is large discrepancy between the calculated contribution and the normalized yields, considering that the 0.3-eV resonance comes from ²⁴¹Am with abundance of 2.29% in the sample. On the other hand, ²³⁹Pu has a resonance at 0.296 eV and ²³⁹Pu is a granddaughter nuclide of ²⁴³Am. It is deduced that the samples contain ²³⁹Pu as impurity. The abundance of ²³⁹Pu was deduced using an area in the neutron energy range from 0.2 to 0.3 eV, abundance of ²⁴¹Am listed in Table 4 and evaluated cross section in JENDL-4.0. The deduced abundance of ²³⁹Pu was 0.8 ± 0.1%.

The uncertainties were derived from the uncertainties due to the area of the resonance. The deduced contribution from ²³⁹Pu and ²⁴⁰Pu is also shown in Figure 12.

3.3. Results and discussion

The normalized yields include influences of the impurities and fission events. Contributions from these influences were calculated using the resolution function of ANNRI and JENDL-4.0. Capture cross sections of ²⁴³Am measured with the three ²⁴³Am samples were derived by subtracting the influences in the neutron energy range of from 10 meV to 100 eV. The derived capture cross sections of ²⁴³Am are shown in Figures 13, 14, and 15.

Figure 13. Derived cross section of the ²⁴³Am measured with the 60-MBq ²⁴³Am sample comparison to the uncertainties of the cross sections (total, statistical uncertainty, uncertainty due to normalization, uncertainty due to dead time correction, uncertainty due to self-shielding and multiple-scattering correction, uncertainty due to neutron flux, uncertainty due to the contribution of fission events, and uncertainty due to impurities).

Figure 14. Derived cross section of the ²⁴³Am measured with the 120-MBq ²⁴³Am sample comparison to the uncertainties of the cross sections (total, statistical uncertainty, uncertainty due to normalization, uncertainty due to dead time correction, uncertainty due to self-shielding and multiple-scattering correction, uncertainty due to neutron flux, uncertainty due to the contribution of fission events, and uncertainty due to impurities).

Figure 15. Derived cross section of the ²⁴³Am measured with the 240-MBq ²⁴³Am sample comparison to the uncertainties of the cross sections (total, statistical uncertainty, uncertainty due to normalization, uncertainty due to dead time correction, uncertainty due to self-shielding and multiple-scattering correction, uncertainty due to neutron flux, uncertainty due to the contribution of fission events, and uncertainty due to impurities).

In this analysis, the statistical uncertainty and the uncertainty due to normalization, the dead time correction, the self-shielding and multiple-scattering correction, the shape of relative neutron flux, the impurity contributions, and the subtraction of fission events were taken into account.

The uncertainties due to impurity subtraction were derived from the uncertainties of the abundances and the uncertainty of the evaluated nuclear data. The uncertainties due to fission events were derived from the uncertainties of the abundances, the uncertainty of the evaluated nuclear data, and the ratio of the sensitivity of fission events to the capture events.

The derived uncertainties are also shown in Figures 13, 14, and 15. The details of these uncertainties at the thermal energy were summarized in Table 6. To reduce the statistical uncertainty, the data shown in Table 6 were averaged over 10 points around neutron energy of 0.0253 eV.

Table 6. Details of uncertainties for the derived capture cross section of ²⁴³Am at neutron energy of 0.0253 eV

Samples	60-MBq sample [b]	120-MBq sample [b]	240-MBq sample [b]
Derived cross section	96	$89.5$	87.7
Statistical uncertainty^a	17	1.8	1.0
Normalization	4	4.4	4.2
Dead time correction	< 1	< 0.1	< 0.1
Self-shielding and multiple-scattering correction	< 1	< 0.1	< 0.1
Relative neutron flux	< 1	0.4	0.4
Fission events subtraction	3	3.1	3.1
Impurities	< 1	0.3	0.3
Total uncertainty	18	5.7	5.4

^aTo reduce the statistical uncertainty, the data were averaged by integrating over 10 points.

Figure 16 shows the obtained neutron capture cross section of ²⁴³Am using the 240-MBq sample with a comparison to evaluated value in JENDL-4.0. The obtained cross section of ²⁴³Am has small peak around 0.305 eV and 0.572 eV. These are the resonance peaks of ²⁴¹Am. It was deduced that the evaluated data of ²⁴¹Am given by JENDL-4.0 are underestimate at the resonances. Obtained cross-section values at neutron energy of 0.0253 eV were found to be $96\pm 18$, $89.5\pm 5.7$, $87.7\pm 5.4$ b. In neutron energy range below 0.1 eV, the present results are agreed with the result by Marie et al. [14], Hatsukawa et al. [15], and Gavrilov et al. [16] as shown in Table 2. However, the results are about 10% larger than the evaluated capture cross section of JENDL-4.0.

Figure 16. Obtained neutron capture cross sections of ²⁴³Am measured with the 240-MBq ²⁴³Am sample and a comparison to evaluated value in JENDL-4.0.

4. Neutron total cross section measurement

4.1. Experimental procedure

For the neutron total cross section measurement, a ⁶Li-glass scintillation detector installed at flight length of 28.7 m was used. The scintillator of the detector was GS-20 (⁶Li enrichment higher than 95%) manufactured by Saint Gobain with dimensions of 50 mm × 50 mm and a thickness of 1 mm. Additionally, to determine background, an enriched ⁷Li-glass detector was used. The scintillator of the ⁷Li-glass detector was GS-30 (⁷Li enrichment higher than 99.99%) manufactured by Saint Gobain with same size and chemical component of the ⁶Li-glass scintillator. The ⁷Li-glass detector was set on the upstream side of the ⁶Li-glass detector. The signals from the photomultipliers (HAMAMATSU:H7195) were analyzed using CAEN V1720 (12bit 250MHz) module with digital pulse processing software, DPP-CI. The module recorded a pulse height and a TOF in ‘event-by-event’ mode.

The 240-MBq ²⁴³Am sample and the dummy sample were used for the measurement. These samples were placed in the most downstream side collimator having a diameter of 9 mm as shown in Figure 1. Flight length from the moderator to the sample position was 24.5 m. To reduce background due to intense gamma-rays from the moderator, lead filters with total thickness of 8.75 cm was placed on the neutron beam line. The total measuring time was about 16 h for the ²⁴³Am sample and about 8 h for the dummy sample. To estimate relative neutron intensities, an Al foil with thickness of 10 $\mu \mathrm{m}$ was set at the sample position of the array of large Ge detectors and prompt gamma-rays from ²⁷Al (n,$\gamma$) reactions were measured with the Ge detectors at the same time.

4.2. Data analysis

Measured pulse height spectra of the ²⁴³Am sample with the ⁶Li-glass and ⁷Li-glass detectors are shown in Figure 17. A peak due to the ⁶Li(n,$\alpha$)³H reactions and double-hit and triple-hit peaks are clearly observed in the spectrum measured by the ⁶Li-glass detector. In the spectrum measured by the ⁷Li-glass detector, a peak due to the ⁶Li(n,$\alpha$)³H reactions are also observed because the ⁷Li-glass scintillator contains slight amount of ⁶Li (0.01%). In this analysis, lower discrimination level at 1900 ch. was applied to derive TOF spectra.

Figure 17. Measured pulse height spectra of the 240-MBq ²⁴³Am sample with the ⁶Li-glass and ⁷Li-glass scintillation detectors.

The gated TOF spectra of the 240MBq ²⁴³Am sample and the dummy sample with the ⁶Li-glass and ⁷Li-glass detectors are shown in Figure 18. Each TOF spectrum was normalized with the number of proton beam shots. Some dips due to ²⁴³Am resonances are clearly observed in the gated TOF spectrum of the ²⁴³Am sample measured by the ⁶Li-glass detector. Because the ⁷Li-glass detector has slight sensitivity to neutron, the strongest resonance of ²⁴³Am is also observed in that measured with the ⁷Li-glass detector.

Figure 18. Gated TOF spectra of the 240-MBq ²⁴³Am sample and the dummy sample with the ⁶Li-glass scintillation detector and the ⁷Li-glass scintillation detector. Each TOF spectrum was normalized with the number proton beam shots.

Dead-time correction was applied to the gated TOF spectra using a fixed dead time of $1264\pm 4$ ns ($316\pm 1$ch.) due to the settings of the CAEN V1720 module. The calculated dead time of the ²⁴³Am sample measurement with the ⁶Li-glass and ⁷Li-glass detectors are shown in Figure 19. The dead time for the ⁶Li-glass detector was less than 2% in the TOF range from 100 $\mu \mathrm{s}$ to 40 ms. The uncertainties of the correction were calculated from the uncertainty of the dead time per event. The influences of the double-hit and triple-hit events were also corrected in this dead-time correction.

Figure 19. Dead time of the 240-MBq ²⁴³Am sample measurements with the ⁶Li-glass detector and the ⁷Li-glass detector.

After the dead-time correction, backgrounds due to frame overlap neutrons and background gamma rays were deduced. The TOF spectra were fitted by an exponential decay plus constant function in the TOF region from 37 to 40 ms, and the result was shifted to the left side by 40 ms. Backgrounds due to gamma-rays were derived from the TOF spectra measured by the ⁷Li-glass detector. Neutron TOF spectra of the ²⁴³Am sample and the dummy sample were derived by subtracting the backgrounds due to the frame overlap neutrons and the gamma-rays.

Relative neutron intensities were derived from the results of the Al foil measurements with the array of large Ge detectors. To eliminate the strong decay gammas from ²⁸Al (1778 keV) and influence of gamma flash, the number of the events whose gamma-ray energies were over 2000 keV and neutron energies were less than 0.1 eV was used. By normalizing the neutron TOF spectra with the deduced relative neutron beam intensities, transmission ratio of the ²⁴³Am sample was obtained. The obtained transmission ratio is shown in Figure 20.

Figure 20. Transmission ratio of the 240-MBq ²⁴³Am sample.

4.3. Results

Total cross section of ²⁴³Am sample was derived in the neutron energy range from 4 meV to 100 eV, using the transmission ratio of the ²⁴³Am sample and the areal density of the ²⁴³Am sample. The areal density was $8.31\pm 0.17\times {10}^3$ atoms/cm² calculated from the activity of the sample ($281.8\pm 0.3$ MBq), half-life of ²⁴³Am ($7357\pm 23$y), and diameter of the sample (10.0 $\pm$ 0.1 mm). The TOF parameters for the neutron energy calibration were derived from a TOF spectrum of a gold sample.

The cross section of ²⁴³Am sample includes influences of the impurities and the difference of Y₂O₃ mass between the ²⁴³Am 240 MBq sample and the dummy sample, as shown in Table 3. Contributions from these materials were calculated using the resolution function of ANNRI and the evaluated total cross sections in JENDL-4.0. Total cross section of ²⁴³Am was calculated by subtracting the contributions from the impurities. The derived total cross section of ²⁴³Am is shown in Figure 21.

Figure 21. Derived total cross section of ²⁴³Am, and uncertainties of the cross sections (total, statistical uncertainty, systematic uncertainty, uncertainty due to neutron intensity, and uncertainty due to impurity subtraction).

In this analysis, the statistical uncertainty, systematic uncertainty, and uncertainty due to neutron intensity and impurities were discussed. The systematic uncertainty derived from the uncertainty of the areal density of the ²⁴³Am 240 MBq sample, i.e. the uncertainties of the sample activity, half-life of ²⁴³Am, and the diameter of the sample. The main component of the systematic uncertainty comes from the uncertainty of the sample diameter. The uncertainty due to neutron intensity derived from statistical uncertainties of the number of measured prompt gamma ray events by the array of large Ge detectors. The value was 0.09%. The uncertainty due to impurity subtraction was derived from the uncertainties of the abundances and the uncertainty of the evaluated nuclear data. The derived uncertainties are also shown in Figure 21. The details of these uncertainties at the thermal energy were summarized in Table 7. The main component of the uncertainty comes from the uncertainty due to neutron intensity at neutron energy of 0.0253 eV and the systematic uncertainty at the strongest resonance (1.356 eV).

Table 7. Details of uncertainties for the derived total cross section of ²⁴³Am at neutron energy of 0.0253 eV

Derived cross section (b)	101
Statistical uncertainty (b)	7.2
Systematic uncertainty (b)	2.3
Uncertainty due to relative neutron intensity (b)	7.4
Uncertainty due to impurity subtraction (b)	1.2
Total uncertainty (b)	11

Obtained cross-section value, at neutron energy of 0.0253 eV, was found to be $101\pm 11$ b. In neutron energy range below 0.1 eV, the present results are about 15% larger than the evaluated total cross section of JENDL-4.0. This under estimate of JENDL-4.0 is also shown in the results of capture cross-section measurement.

5. Conclusion

This work describes neutron capture and total cross-section measurements of the nucleus ²⁴³Am carried out at the ANNRI in J-PARC.

The neutron capture cross section of ²⁴³Am was measured with the array of Ge detectors using the three ²⁴³Am samples with different activities. Absolute capture cross section at the 1.356 eV resonance was derived by taking ratios of the capture yields. The neutron capture cross sections were determined in the neutron energy range from 10 meV to 100 eV by normalizing the relative yield to the derived absolute cross section at the 1.356-eV resonance. The determined capture cross sections at neutron energy of 0.0253 eV were found to be $87.7\pm 5.4$ b. In the neutron energy range below 0.1 eV, the present results are about 10% higher compared to the evaluated capture cross section of JENDL-4.0.

The neutron total cross section of ²⁴³Am was measured with the ⁶Li and ⁷Li-enriched detectors. By using the enriched ⁷Li-glass scintillator with same size and chemical component of the enriched ⁶Li-glass scintillator, the backgrounds were evaluated. Total cross section of ²⁴³Am was derived in the neutron energy range from 4 meV to 100 eV. The determined cross section at neutron energy of 0.0253 eV was 101 ± 11 b. In neutron energy range below 0.1 eV, the present results are about 15% larger than the evaluated total cross section of JENDL-4.0.

Acknowledgments

Present study includes the result of ‘Research and Development for accuracy improvement of neutron nuclear data on minor actinides’ entrusted to the Japan Atomic Energy Agency by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).

Disclosure statement

No potential conflict of interest was reported by the authors.