Neutron-induced Reaction Rates Research Articles

NG-TRAP: Measuring neutron capture cross-sections of short-lived fission fragments

We lack significant nuclear physics input to understand the rapid-neutron capture (r-)process fully. The r-process is the source of half the elements heavier than iron and the only way to produce the long-lived actinides we find on earth. This process’s key nuclear physics inputs are nuclear masses, cross-sections of (n,γ) and (γ,n), and decay half-lives and branching ratios of neutron-rich isotopes. However, there is currently no method to directly measure neutron-induced reaction rates on short-lived nuclides, so there is no experimental data for the primary nuclear reaction that drives the r-process. We show here a conceptual design of a novel approach to access this information experimentally. The idea is to form a target of short-lived isotopes by confining them as ions in a radio-frequency (RF) trap. Next, they are irradiated with an intense neutron flux, and the reaction products are identified by mass spectrometry. The chosen method is a two-stage process in the presence of high neutron fluxes. The first process is neutron-induced fission in a thin actinide foil to create fission fragments. These fragments are slowed down in a cryogenic stopping cell before being filtered through a radio frequency quadrupole (RFQ) system. The RFQ system selects fission fragments of a specific atomic mass number A and confines them to a small volume in an RF trap, where they are irradiated for a second time in a controlled manner. The resultant A+1 isotopes are mass-selectively transported to a multiple-reflection time-of-flight mass spectrometer, where the reaction products are identified and counted.

Experimental study on tritium breeding in water-cooled ceramic breeder blanket mock-up under D–T neutron irradiation conditions

The accurate assessment of tritium breeding parameters within fusion blankets is crucial for future magnet-confined fusion machines to realize fuel self-sufficiency. Such an assessment can be conducted using the simulation approach with nuclear data of high-fidelity and, most importantly, validated against experimental data. In this paper, we report neutronics experiment studies carried out on a mock-up of the water-cooled ceramic breeder blanket of the China fusion engineering test reactor (CFETR), irradiated at a deuterium–tritium (D–T) neutron source. The mock-up’s nuclear responses to 14 MeV neutrons, including tritium production rates (TPR) and neutron-induced reaction rates, are simulated and validated by experimental results. Redundant measurement techniques are used, including Li2TiO3 pellets for offline TPR measurements and the developed lithium glass detector with a significantly reduced size for online TPR measurements. The validation of the TPR value is complemented by the experimental evaluation of the neutron-induced reaction rates for Au and Zr foils. All experimental results are analyzed using the MCNP-4C code and FENDL-3.0 nuclear data library. The source term for the Monte Carlo simulation is built using a newly-developed method based upon the modeling of the depth profiling of tritium in the tritiated target. The experiments are in good agreement with the simulations; ratios of the calculation to experimental results (C/E) on TPR are found to be 0.97–1.08. The influence of the first wall tungsten armor on the mock-up nuclear responses is also studied.

Measurement of thick target neutron yield at 180°for a mercury target induced by 3-GeV protons

A thick target neutron yield (TTNY) for a mercury target at an angle of 180°from the incident beam direction was measured with the time-of-flight method using a 3-GeV proton beam at the Japan Proton Accelerator Research Complex (J-PARC). Comparison of the experimental results with a Monte Carlo particle transport simulation by the Particle and Heavy Ion Transport code System (PHITS) shows that the default spallation model of PHITS, the Liège intranuclear cascade model version 4.6 coupled with the generalized evaporation model (INCL4.6/GEM) reproduces the measured TTNY fairly well, whereas apparent overestimation is observed for the former default model, the Bertini intranuclear cascade model coupled with GEM (Bertini/GEM). These trends are consistent with the experimental results of neutron-induced reaction rates obtained using indium and niobium activation foils.

On 54Fe neutron cross section importance in iron

The presented paper aims to evaluate the importance of 54Fe XS in iron by means of measuring the reaction rates of the selected reactions on 54Fe and measuring a fast neutron leakage spectra from the iron sphere of 100cm in diameter by a stilbene scintillation detector with subsequent XS sensitivity analysis. The reactions involved in the study were 54Fe(n,p) and 54Fe(n,α). Measured neutron induced reaction rates in 54Fe are compared with calculated ones in different nuclear data libraries. We show that there are notable discrepancies in 54Fe(n,α) reaction. The results of the leakage spectra differ significantly in various libraries, library ENDF/B-VII.1 in region 3.5–7.0MeV gives relatively good agreement. CIELO library underestimates the result; however JEFF-3.2 overestimates results., 252Cf with the emission rate of 9.53E8 n/s was used as a neutron source for all experiments involved.

Measurement of reaction rates for different neutron induced reactions in 27Al

The presented paper aims to compare various measured neutron induced reaction rates in Aluminium with computed ones in different nuclear data libraries. A 252Cf neutron source with emission rate of 9.53E8 n/s was used. Reactions involved in the study were 27Al(n,g), 27Al (n,p) and 27Al (n,α).

Study of secondary neutron interactions with 232Th, 129I, and 127I nuclei with the uranium assembly “QUINTA” at 2, 4, and 8 GeV deuteron beams of the JINR Nuclotron accelerator

The natural uranium assembly, “QUINTA”, was irradiated with 2, 4, and 8GeV deuterons. The 232Th, 127I, and 129I samples have been exposed to secondary neutrons produced in the assembly at a 20-cm radial distance from the deuteron beam axis. The spectra of gamma rays emitted by the activated 232Th, 127I, and 129I samples have been analyzed and several tens of product nuclei have been identified. For each of those products, neutron-induced reaction rates have been determined. The transmutation power for the 129I samples is estimated. Experimental results were compared to those calculated with well-known stochastic and deterministic codes.

Neutron flux spectra investigations in breeding blanket assembly containing lithium titanate and Lead irradiated with DT neutrons

A benchmark experiment was performed to obtain the neutron spectra in a breeding blanket assembly consisting of lithium titanate and Lead using multiple foil activation technique with unfolding code SAND-II-SNL. The purpose of the experiment was to validate the calculated neutron spectra and neutron induced reaction rates with MCNP-4A code and FENDL-2.1 & FENDL-3.0 transport libraries in the layered geometry of lithium titanate and Lead. Measured reaction rates in different activation foils were compared with MCNP-IRDFF-1.05 and FISPACT-EAF2007 calculated values. In general, FENDL-3.0 calculated reaction rates are closer to the measured reaction rates than FENDL-2.1 calculated rates. Calculated Lead multiplication is improved from FENDL-2.1 to FENDL-3.0 up to 2%. Unfolded neutron spectra in the multiplier zones are slightly underestimated by calculations in the secondary neutron emission range of Pb(n, xn) reactions indicating the need of improvement in the (n, 2n) cross-section of Lead isotopes. The observed underestimation in the Lead multiplication puts the tritium production calculations of breeding blankets using Lead as multiplier on conservative side. The inelastic scattering range of Lead isotopes is overestimated by calculations as large as 80% in one of the breeder zone. This observation requires improvement in the inelastic scattering cross-sections of Lead isotopes in FENDL evaluations.

Neutron-induced reaction rates for the r-process

Neutron-induced reaction rates for the formation of heavy and superheavy nuclei in the astrophysical r-process are presented. Neutron-induced reactions, including fission and neutron capture, at astrophysical temperatures are treated within the framework of the statistical model and calculated for targets with the atomic number 84 < Z < 118, using different mass and fission barrier predictions. The dependence of heavy element yields from nuclear explosions on the target nuclei, nuclear reactions, and nuclear models is discussed.

Neutron-induced astrophysical reaction rates for translead nuclei

Neutron-induced reaction rates, including fission, are calculated in the temperature range 1.d8 <T (K) < 1.d10 within the framework of the statistical model for targets with atomic number 83 < Z < 119 (from Po to Uuo) from the neutron to the proton drip-line. Four sets of rates have been calculated, utilizing - where possible - consistent nuclear data for neutron separation energies and fission barriers from Thomas-Fermi (TF), Extended Thomas-Fermi plus Strutinsky Integral (ETFSI), Finite-Range Droplet Model (FRDM) and Hartree-Fock-Bogolyubov (HFB) predictions. Tables of calculated values as well as analytic seven parameter fits in the standard REACLIB format are supplied. We also discuss the sensitivity of the rates to the input, aiming at a better understanding of the uncertainties introduced by the nuclear input.

Measurements with the high flux lead slowing-down spectrometer at LANL

A Lead Slowing-Down Spectrometer (LSDS) was recently installed at LANL [D. Rochman, R.C. Haight, J.M. O’Donnell, A. Michaudon, S.A. Wender, D.J. Vieira, E.M. Bond, T.A. Bredeweg, A. Kronenberg, J.B. Wilhelmy, T. Ethvignot, T. Granier, M. Petit, Y. Danon, Characteristics of a lead slowing-down spectrometer coupled to the LANSCE accelerator, Nucl. Instr. and Meth. A 550 (2005) 397]. The LSDS is comprised of a cube of pure lead 1.2m on the side, with a spallation pulsed neutron source in its center. The LSDS is driven by 800MeV protons with a time-averaged current of up to 1μA, pulse widths of 0.05–0.25μs and a repetition rate of 20–40Hz. Spallation neutrons are created by directing the proton beam into an air-cooled tungsten target in the center of the lead cube. The neutrons slow down by scattering interactions with the lead and thus enable measurements of neutron-induced reaction rates as a function of the slowing-down time, which correlates to neutron energy. The advantage of an LSDS as a neutron spectrometer is that the neutron flux is 3–4 orders of magnitude higher than a standard time-of-flight experiment at the equivalent flight path, 5.6m. The effective energy range is 0.1eV to 100keV with a typical energy resolution of 30% from 1eV to 10keV. The average neutron flux between 1 and 10keV is about 1.7×109n/cm2/s/μA. This high flux makes the LSDS an important tool for neutron-induced cross section measurements of ultra-small samples (nanograms) or of samples with very low cross sections. The LSDS at LANL was initially built in order to measure the fission cross section of the short-lived metastable isotope of U-235, however it can also be used to measure (n,α) and (n,p) reactions. Fission cross section measurements were made with samples of 235U, 236U, 238U and 239Pu. The smallest sample measured was 10ng of 239Pu. Measurement of (n,α) cross section with 760ng of Li-6 was also demonstrated. Possible future cross section measurements include fission and (n,p) and (n,α) reaction in radioactive samples.

Neutron flux parameters at irradiation positions in the new research reactor FRM-II

The new research reactor FRM-II in Garching, Germany, was at full power 20 MW for the first time on 24th August, 2004. Since then, highly thermalized neutrons are available also for neutron activation analysis (NAA). In this report, all essential neutron flux parameters needed to calculate neutron induced reaction rates based on the Høgdahl or Westcott convention are presented for all irradiation positions in this reactor.

Deterministic and Monte Carlo Neutron Transport Calculations of the Dounreay Fast Breeder Reactor

A homogenized whole-reactor cylindrical model of the Dounreay Fast Reactor has been constructed using both deterministic and Monte Carlo codes to determine neutron flux distributions inside the core and at various out-of-core components. The principal aim is to predict neutron-induced activation levels using both methods and make comparisons against the measured thermal reaction rates. Neutron transport calculations have been performed for a fixed source using a spatially lumped fission neutron distribution, which has been derived from measurements. The deterministic code used is based on the finite element approximation to the multigroup second-order even-parity neutron transport equation, which is implemented in the EVENT code. The Monte Carlo solutions were obtained using the MCNP4C code, in which neutron cross sections are represented in pointwise (or continuous) form. We have compared neutron spectra at various locations not only to show differences between using multigroup deterministic and continuous energy (point nuclear data) Monte Carlo methods but also to assess neutron-induced activation levels calculated using the spectra obtained from both methods. Results were also compared against experiments that were carried out to determine neutron-induced reaction rates. To determine activation levels, we employed the European Activation Code System FISPACT. We have found that the neutron spectra calculated at various in-core and out-of-core components show some differences, which mainly reflect the use of multigroup and point energy nuclear data libraries and methods employed, but these differences have not resulted in large errors on the calculated activation levels of materials that are important (such as steel components) for decommissioning studies of the reactor. The agreement of calculated reaction rates of thermal neutron detectors such as the 55Mn(n,γ)56Mn against measurements was satisfactory.

Neutron and photon flux spectra in a mock-up of the ITER shielding system

Abstract A neutronics mock-up of the ITER inboard shielding system including first wall, shield-blanket, vacuum vessel and magnetic field coils, was investigated in order to validate calculational tools used for the ITER design and to verify the nuclear performance of the shield by measuring neutron and photon flux spectra at two positions inside the bulk assembly. The measured spectra are compared with results from calculations using the Monte Carlo code MCNP-4A and the data library FENDL-1. Integral values such as neutron-induced reaction rates and gamma-heating are calculated on the basis of the experimentally obtained flux spectra, and are compared with independent measurements of these data. Helium and hydrogen production rates, as well as the radiation damage rate, are determined again by using the measured flux spectra.

Thin metal encapsulation of fission foils and oxide discs by resistance welding

The Interlaboratory LMFBR Reaction Rate (ILRR) Program was established in 1971 by the U.S. Atomic Energy Commission, Division of Reactor Research and Development. The objective of this program was the development of the capability to precisely measure neutron-induced reaction rates of LMFBR fuels and related materials. The Isotope Research Materials Laboratory of the Oak Ridge National Laboratory has been actively involved in development and fabrication of various types of dosimeters to meet the demands of this program since its inception. One type of dosimeter utilized in this program has been thin metallic and oxide discs of uranium, plutonium, and neptunium isotopes. Because of the extreme reactivity and radiotoxicity of these materials, it has been necessary to fabricate these dosimeters with a welded closure. The stringent dimensional specifications and the extreme thinness of the container materials precluded the use of conventional welding techniques. This report describes the equipment and procedures employed to attain the welded closures required to hermetically seal these thin metal containers.

The accurate fission cross-section of 235U from 0·002 to 0·15 eV and its reference value at 2200 m/s

A new precise measurement of the 2200 m/s fission cross-section of 235U was performed with a slow chopper at the BR2 reactor (energy-range from 0·002 to 0·15 eV). Simultaneous recordings of pulse-height and time-of-flight spectra of successively compared 235UF 4-layers and “standard” elemental boron-layers were made. The detection of the 10B( n, α) 7Li- and 235U( n, f) reaction fragments is made with excellent pulse-height resolution in low geometry with a surface-barrier detector. The absolute 10B contents were determined from direct weighing in vacuum combined with careful chemical and isotopic analysis of witness foils. The relative 10B contents of the standard boron-foils were checked by their neutron induced reaction rates in the geometry of the fission experiment. The amount of uranium on the fissile foils is determined by precise low geometry α-counting making use of the most recent and accurate α half-life values. The fission foils were also checked by α- and fission fragment counting in the geometry of the experiment. The fission cross-section was calculated in the energy-range from 0·002 to 0·15 eV and the absolute value deduced at 2200 m/s was 587·6 ± 2·6 barn. The value at 25 × 10 −3 eV is 591·8 ± 2·7 barn. These fission cross-section values were used to calculate the Westcott- g φ -factor for a 20·44°C Maxwellian-distribution with as result: g f = 0·9780 ± 0·0010.