JENDL Fusion File Research Articles

06/00089 Importance of the MUSE experiments foremerging ADS concepts from the nuclear data viewpoint: Plaschy, M. et al. Annals of Nuclear Energy, 2005, 32, (8), 843–856

Fusion integral benchmark experiments have been carried out for copper and tungsten at FNS of JAERI. Leakage neutron spectra were measured with an NE213 detector by the TOF method to validate evaluated nuclear data files of JENDL-3.2, JENDL-fusion file, ENDF/B-VI, FENDL/E-1.0 and FENDL/E-2.0. From the result of comparison between experiments and calculations by MCNP-4B, below 1 MeV disagreement was observed for copper and it became remarkable with increase of the thickness. It was therefore, pointed out that there might be a problem in the evaluated (n, 2n) neutron spectrum for all the nuclear data libraries. For tungsten, the calculated result with JENDL-fusion file and FENDL/E-2.0 overpredicted the experimental values around 3–7 MeV. Near 10 MeV, underestimation of ENDF/B-VI was remarkably pronounced. For the whole energy range, JENDL-3.2 was preferable. However, there was a serious problem left that discrepancy between JENDL-fusion file and JENDL-3.2 was very large.

Fusion Neutronics Benchmark Experiment on Structural and Advanced Blanket Materials - Leakage Gamma-ray Spectrum Measurement -

Benchmark experiments have been carried out to validate the nuclear data files of JENDL-3.2, JENDL fusion file, ENDF/B-VI, FENDL/E-1.0 and FENDL/E-2.0 for blanket materials of LiA1O2, Li2TiO3 and Li2ZrO3 and structural materials, i.e., C, V, Fe, SS-316, Cu, Pb and W. Some discrepancies between measured and calculated spectra were observed. However, the C/E values for the energy multiplied integral spectrum show that all of the nuclear data files were fairly reliable.

Fusion Neutronics Benchmark Experiment on Structural and Advanced Blanket Materials - Leakage Neutron Spectrum Measurement -

Fusion neutronics benchmark experiments for copper and tungsten and advanced breeder materials of LiAlO2, Li2TiO3 and Li2ZrO3 have been conducted at FNS of JAERI to validate five nuclear data files, i.e., JENDL-3.2, JENDL Fusion File, ENDF/B-VI, FENDL/E-1.0 and FENDL/E-2.0. From the result, all the nuclear data files were confirmed to be fairly reliable with respect to the prediction of neutron spectrum. However, some problems were pointed out. Especially for Zr, the discrepancy between the experimental and calculated spectra is remarkably large.

Neutronics Experiment of 6Li-enriched Breeding Blanket with Li2TiO3/Be/F82H Assembly Using D-T Neutrons

To improve the nuclear performance of fusion breeding blankets, the application of thermal type blankets with 6Li-enriched ceramics and low-activation ferritic steel is suggested. However, appropriate neutronics experiments to investigate the nuclear performance of such blankets have never been carried out. Therefore, we have done blanket neutronics experiments with stratified 95-% enriched 6Li2TiO3, stainless steel F82H and beryllium blocks using JAERI’s Fusion Neutronics Source (FNS) and verified the prediction accuracy of the tritium production rate and the production rate of selected isotopes in F82H in with the Monte Carlo code MCNP-4B and JENDL Fusion File, JENDL-3.2 and ENDF/B-VI data. Calculated tritium and 51Cr production rates were in good agreement with the measured values within the experimental error. However, discrepancies were found for 187W and 56Mn production rates.

Recent Measurements of Nuclear Data and Integral Tests for Fusion Reactor Application

Differential and integral experiments including measurements of neutron, gamma-ray and charged particle emission DDXs, (n,2n) reaction cross section and leakage neutron and gamma-ray spectra from slab assemblies have been carried out at OKTAVIAN of Osaka University and FNS of JAERI. The measured data were utilized for benchmarking of five nuclear data files of JENDL-3.2, JENDL-fusion file, ENDF/B-VI, FENDL/E-1.0 and FENDL/E-2.0. In the present paper, their recent results are summarized.

JENDL Fusion File 99

The double-differential cross sections (DDXs) of secondary neutrons have been evaluated for 79 isotopes and 13 natural elements ranging from H to Bi to improve the accuracy of predictions for the neutronics calculations in the D-T thermonuclear fusion applications. The data given in JENDL-3.1, which was the newest version of JENDL general purpose file when this project was initiated, was combined with new calculations based on the optical model, DWBA, pre-equilibrium and multi-step statistical models, and the DDX data were generated based on various kinds of systematics for medium-mass nuclei. Different methods were employed for light nuclei to which the above method could not be applied. In addition, the DDXs for emission of charged particles (p, d, t, 3He and α-particle) were given for 2H, 9Be and elements heavier or equal to F. The present results give an overall good description of the measured DDX data of both the neutron and charged particles emission channels. The data were compiled in ENDF-6 format, and released in 1999 as a special purpose file of JENDL family, namely, JENDL Fusion File 99.

Neutron–nuclear data benchmark for copper and tungsten by slab assembly transmission experiments with DT neutrons

Fusion integral benchmark experiments have been carried out for copper and tungsten at FNS of JAERI. Leakage neutron spectra were measured with an NE213 detector by the TOF method to validate evaluated nuclear data files of JENDL-3.2, JENDL-fusion file, ENDF/B-VI, FENDL/E-1.0 and FENDL/E-2.0. From the result of comparison between experiments and calculations by MCNP-4B, below 1 MeV disagreement was observed for copper and it became remarkable with increase of the thickness. It was therefore, pointed out that there might be a problem in the evaluated (n, 2n) neutron spectrum for all the nuclear data libraries. For tungsten, the calculated result with JENDL-fusion file and FENDL/E-2.0 overpredicted the experimental values around 3–7 MeV. Near 10 MeV, underestimation of ENDF/B-VI was remarkably pronounced. For the whole energy range, JENDL-3.2 was preferable. However, there was a serious problem left that discrepancy between JENDL-fusion file and JENDL-3.2 was very large.

Benchmark experiment on silicon carbide with D–T neutrons and validation of nuclear data libraries

Silicon carbide (SiC) is one of the low activation structural materials for fusion reactors. A benchmark experiment on SiC was conducted at the deuterium–tritium (D–T) neutron source facility FNS in JAERI, and the validity of the cross section data for SiC was investigated. An experimental assembly made of sintered SiC (457×457 mm, and 711 mm in thickness) was bombarded by D–T neutrons, and neutron spectra, gamma-ray spectra, dosimetry reaction rates and gamma-ray heating rates were measured in the assembly. Computational analyses of the experiment were performed by Monte Carlo transport calculations with the MCNP-4B code and several recent evaluated nuclear data files. The following findings were pointed out. The calculations with JENDL Fusion File and JENDL-3.2 predicted adequately the neutron fluxes from 14 MeV to 0.3 eV as well as the gamma-ray fluxes at any positions up to ∼600 mm depth in the SiC assembly. Results for the calculation with FENDL/E-2.0 were similar to those with the two JENDL calculations, but the neutron fluxes in the energy range from 1 to 14 MeV were calculated slightly smaller. The cross section data for silicon in FENDL/E-1.0 were found inadequate for nuclear design calculations.

Overview of straight duct streaming experiments for ITER

Two straight duct streaming experiments for international thermonuclear experimental reactor (ITER) were performed by using the intense deuterium–tritium (D–T) neutron source at Japan Atomic Energy Research Institute (JAERI) in order to investigate the influence of streaming through ducts on radiation dose and to validate the accuracy of design calculations. One was a small duct streaming experiment for diagnostic ports, the other was a large duct streaming experiment for large ports such as neutral beam injector (NBI) port. The ducts increased the neutron flux above 10 MeV by 10 6–10 7 times at the ends of the ducts, while the increase of neutron flux below 10 MeV and γ-ray by the ducts was less than a few hundred times. These experiments were analyzed by the MCNP-4A code with three nuclear data libraries: FENDL/E-2.0, -1.0 and JENDL Fusion File. Though the ducts made complicated neutron and γ-ray flux distributions in a cavity region, most of the calculated results agreed within ±40% with the experimental data. It was demonstrated that the MCNP calculations with the nuclear data libraries were accurate enough for predicting nuclear design parameters of ITER even for various duct geometries as far as they were modeled precisely.

Benchmark experiment on LiAlO 2, Li 2TiO 3 and Li 2ZrO 3 assemblies with D–T neutrons–leakage neutron spectrum measurement

For the three materials LiAlO 2, Li 2TiO 3 and Li 2ZrO 3, known as advanced breeder materials for a fusion reactor, D–T neutron benchmark experiments have been conducted at fusion neutronics source (FNS) of Japan Atomic Energy Research Institute (JAERI), Japan, to validate five nuclear data files, i.e. JENDL-3.2, JENDL fusion file, ENDF/B-VI, FENDL/E-1.0 and FENDL/E-2.0. From the result, all the nuclear data files were confirmed to be fairly reliable with respect to the prediction of neutron spectrum in the use of LiAlO 2 and Li 2TiO 3. However, it was recommended that the inelastic scattering neutron spectra of oxygen, lithium and aluminium in the five nuclear data files had to be re-examined. As for zirconium, it was pointed out that the natZr(n, 2n) cross section could be overestimated for the five nuclear data files.

Measurement and Analysis of Leakage Neutron Spectra from Spherical Assemblies of Chromium, Manganese and Copper with 14 MeV Neutrons

In order to make benchmark validation of the nuclear data for Cr, Mn and Cu in existing modern evaluated nuclear data files, leakage neutron spectra in the energy range from 0.1 to 15 MeV from spherical sample piles were measured by a time-of-flight technique using intense 14 MeV neutron source, OKTAVIAN. To make a qualitative understanding of the dependence of the spectra on individual partial cross sections, a trial calculation was performed by Monte Carlo neutron transport code MCNP-4A along with several kinds of modified cross sections for Mn. The measured spectra were compared with those by theoretical calculation using MCNP-4A with the evaluated nuclear data files. The comparison was made by the spectrum shape and by the C/E values in four energy regions. The calculations for Cr, Mn and Cu using JENDL-3.2, JENDL Fusion File, FENDL-1.0 and EFF-2 give satisfactory predictions with a few exceptions. By checking the result of the trial calculation, it is considered that these disagreement is caused by improper evaluation either of inelastic scattering cross section values and/or energy distribution of secondary neutrons.

Measurement of the Neutron Spectrum in the 14MeV Neutron Driven Vanadium-Alloy Assembly by the Multifoil Activation Method

A neutronics benchmark experiment on vanadium-alloy was conducted by using the D-T neutron source facility of FNS/ JAERI. The neutron spectra in the vanadium-alloy assembly which was bombarded by D-T neutrons were measured by the multifoil activation - adjustment method. Reaction rates of 20 dosimetry reaction were measured. Neutron spectra were calculated by the MCNP-4A code with the JENDL Fusion File and FENDL/E-1.0 libraries. These calculated neutron spectra were used as initial guesses and adjusted by the SAND-II code with the measured reaction rates. Cross sections of the FENDL/A-2.0 library were used as detector response functions for the adjustment. The obtained adjusted spectra are consistent with the results of the on-line spectrum measurement. The evaluated cross section data of vanadium-alloy were validated on the comparisons between the calculated initial spectra and the adjusted spectra.

Investigation on Prediction Capability of Nuclear Design Parameters for Gap Configuration in ITER through Analysis of the FNS Gap Streaming Experiment

As an R&D Task of shielding neutronics experiment under the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER), streaming experiments with simulating a gap configuration formed by two neighboring blanket modules of ITER were carried out at the FNS facility. In this work, prediction capability of various nuclear design parameters was investigated through analysis of the experiments. The Monte Carlo transport calculation code MCNP-4A and the FENDL/E-1.0 and JENDL Fusion File cross section data libraries were used for the analysis with detailed modeling of the experimental conditions. As a result, all the measured quantities were reproduced within about ± 30 % by the calculations. It was concluded that these calculation tools were capable of predicting nuclear design parameters, such as helium production rates at connection legs of blanket modules to the back plate and nuclear responses in toroidal field coils, with uncertainty of ± 30% for the geometry where gap streaming effect was significant.

Measurement and Analysis of Leakage Neutron Spectra from Spherical Assemblies of Chromium, Manganese and Copper with 14 MeV Neutrons.

In order to make benchmark validation of the nuclear data for Cr, Mn and Cu in existing modern evaluated nuclear data files, leakage neutron spectra in the energy range from 0.1 to 15 MeV from spherical sample piles were measured by a time-of-flight technique using intense 14 MeV neutron source, OKTAVIAN. To make a qualitative understanding of the dependence of the spectra on individual partial cross sections, a trial calculation was performed by Monte Carlo neutron transport code MCNP-4A along with several kinds of modified cross sections for Mn. The measured spectra were compared with those by theoretical calculation using MCNP-4A with the evaluated nuclear data files. The comparison was made by the spectrum shape and by the C/E values in four energy regions. The calculations for Cr, Mn and Cu using JENDL-3.2, JENDL Fusion File, FENDL-1.0 and EFF-2 give satisfactory predictions with a few exceptions. By checking the result of the trial calculation, it is considered that these disagreement is caused by improper evaluation either of inelastic scattering cross section values and/or energy distribution of secondary neutrons.

Measurements of Double-Differential Cross Sections of Charged-Particle Emission Reactions for Several Structural Elements of Fusion Power Reactors by 14.1-MeV Incident Neutrons

A two-dimensional energy and time-of-flight charged-particle spectrometer has been developed and used to measure the double-differential cross-section (DDX) data of (n, xp) and (n, xα) reactions for several elements with 14.1-MeV incident neutrons at OKTAVIAN, the Intense 14-MeV Neutron Source Facility of Osaka University. The DDX data of the 51V(n, xp), 51V(n, xα), natFe(n, xp), natFe(n, xα), 59Co(n, xp), 59Co(n, xα), natNi(n, xα), natCu(n, xα), 93Nb(n, xp), 93Nb(n, xα), and natMo(n, xp) reactions are measured. The angle-integrated energy differential cross-section (EDX) data were deduced from the measured DDX data and compared with other experimental results [except for the 59Co(n, xp) reaction] and evaluated nuclear data of JENDL fusion file (JENDL-FF). A comparison was also done with the ENDF/B-VI for the total reaction cross sections of all measured reactions except for the natMo(n, xp) reaction and the EDX of the natNi(n, xα) and natCu(n, xα) reactions. The theoretical calculations were done by using the SINCROS-II code. The measured data agreed fairly well with other data for almost all the reactions. The JENDL-FF and SINCROS-II data underestimate the measured EDX data for the reactions of 93Nb(n, xα) and natMo(n, xp). For the natFe(n, xp), natFe(n, xα), 59Co(n, xα), and natNi(n, xα) reactions, smaller data are given than other data, i.e, other experimental data, JENDL-FF, and ENDF/B-VI. The SINCROS-II code can reproduce well for both the proton and alpha-particle emission cross-section values.

Measurements of Double-Differential Neutron Emission Cross Sections of 6Li, 7Li and 9Be for 11.5 MeV and 18.0 MeV Neutrons

Double-differential neutron emission cross sections (DDXs) of 6Li, 7Li and 9Be were measured for 18.0 MeV and 11.5 MeV incident neutrons produced by the T(d, n) and 15N(d, n) reactions respectively, using the Tohoku University Dynamitron time-of-flight (TOF) spectrometer. The data were obtained at 13 laboratory angles, and angular-differential cross sections (ADXs) of elastic and inelastic scattering neutrons were derived from the DDXs. For 11.5 MeV neutrons, we obtained the neutron emission spectra over the secondary neutron energies by newly employing the double TOF method as well as the conventional one. In the measurements at 18.0 MeV, we achieved better energy resolution than in our previous studies by using a neutron detector that has a larger solid angle and a thinner tritium target. The experimental results of DDXs and ADXs were compared with our previous results and the evaluated data given in JENDL-3.2, JENDL Fusion File and ENDF/B-VI. It is found that the JENDL data reproduce the experimental ones very well.

Some Comments for Cross Section Data of Iron around 15 MeV

An analysis of benchmark experiment on iron for D-T neutrons with JENDL Fusion File and FENDL/E-1.1 suggested that neutron flux above 10 MeV in iron, was underestimated monotonously with depth. Reasons of this underestimation were investigated through various analyses by DORT3.1 with JENDL Fusion Füe, FENDL/E-1.1 and FENDL/E-2.0. The followings for evaluated cross section data on iron around 15 MeV were considered to be possible origins of underestimation of neutron flux above 10 MeV.1. JENDL Fusion File: Elastic scattering cross sections for forward angles were smaller. Angle-integrated cross section data of (n,2n) and (n,np) reactions were larger.2. FENDL/E-1.1: Elastic scattering cross sections for forward angles were smaller.3. FENDL/E-2.0: Angle-integrated cross section data of inelastic scattering and (n,np) reaction were larger.

Benchmark Experiment on Vanadium Assembly with D-T Neutrons -Leakage Neutron Spectrum Measurement-

Benchmark experiments on vanadium and vanadium alloy with D-T neutrons have been done at two angles, 0 degrees and 24.9 degrees, using the slab geometry and the time-of-flight (TOF) method. Data were collected for neutron energies ranging from 50 keV to 15 MeV. For vanadium, measurements were made for three slab thicknesses, i.e., 50.8 mm, 1524 mm, and 254 mm, whereas for the vanadium alloy, measurements were made only for 101.6-mm thickness. The measured neutron spectra were compared with MCNP-4A calculations using evaluated nuclear data from the JENDL-3.2, JENDL Fusion-File(IENDL-FF), FENDL/E-1.0 and European Fusion File veraon-3(EFF-3) libraries. The calculated data show reasonable agreement with the measurement, however, some differences are worth noting. Calculations for a slab thickness of 50.8 mm over the energy range from 0.05 to 0.1 MeV underestimate the measurements by about 40% at an angle of 24.9 degrees, while calculations for the energy range from 0.1 to 1.0 MeV, overestimate the measurements by about 40% at an angle of 0 degrees. Calculations made using the JENDL-FF library show good agreement with measurements for energies greater than 11 MeV. Calculations made using the FENDL/E-1.0 library give smaller results than any of the other three libraries in the energy range from 5 to 11 MeV.

Benchmark experiment on bulk shield of SS316/water with simulated superconducting magnet

Abstract A bulk shielding experiment was performed at the D-T neutron source FNS, using the assembly simulating shield blanket, vacuum vessel and toroidal field coil (superconducting magnet) of ITER. The objective was to investigate the influence on nuclear parameters due to the nuclei involved in the toroidal field coil and the effect of B4C/Pb. The analysis was performed by MCNP4-A and the DOT3.5 (P5S16 approximation, 175-n and 42-γ groups with and without self-shielding correction) with JENDL Fusion File (JENDL-FF) and FENDL/E-1.1. All of the MCNP calculations and DOT calculations with self-shielding correction agreed within ± 40% with the measurements. It was found that the influence of self-shielding mainly due to copper and tantalum in the DOT calculations was large for neutrons less than 10 keV. It is concluded that the uncertainty of nuclear designs by MCNP-4A and DOT3.5 with JENDL-FF and FENDL/E-1.1 on the bulk shield of ITER is approximately 40% in the toroidal coil region. The analysis suggested that it was quite important to deal with heavy nuclei of small quantity such as tantalum precisely in the shielding design.

Nuclear heating measurements for SS-316, copper, graphite, tungsten, chromium, beryllium in a copper centered assembly bombarded with 14MeV neutrons and analysis

Under ITER/EDA R&D Task T-218, an experiment on nuclear heating was conducted at the Fusion Neutronics Source Facility (FNS) in JAERI in order to provide experimental data of nuclear heating for structural materials and to validate the data and methods relevant to ITER design calculations. Probe materials, SS-316, Cu, W, Zr, Be, Cr and graphite were tested in a copper centered experimental assembly bombarded with 14 MeV neutrons. A calorimetric method and TLDs were employed for total heating and γ-heating measurements, respectively. The measurements were carried out along a central axis of a copper centered experimental assembly. The total heating rate in the probe material was derived from a temperature rise measured by a thermistor attached to the probe. MCNP4A with JENDL-3.2, JENDL-Fusion File and FENDL-1 libraries were used for experimental analyses. Nuclear heating was derived with the use of KERMA factors based on JENDL-3.2 and FENDL-1. The adequacy of the KERMA data is discussed based on nuclear heating ratios of calculation to experiment (C/E).