Thermal Neutron Spectrum Research Articles

Assessment of MICROX-2 Code with New ENDF/B-VII Release 0 Master Libraries

A lattice code, MICROX-2, was assessed for its neutronics calculation performance with new cross-section libraries. First, the new cross-section libraries were generated based on ENDF/B-VII release 0. A total of 386 nuclides were processed, including 10 thermal scattering nuclides. The updated NJOY system and MICROR code were used to process nuclear data and convert them into the MICROX-2 library format. The energy group structure of the new library was optimized for both the thermal and fast neutron spectrum systems based on the Contributon and Pointwise Cross Section Driven (CPXSD) method, resulting in a total of 1173 energy groups. Second, a series of pin-cell–level benchmark calculations was performed against experimental measurements and numerical calculations performed by the deterministic and Monte Carlo codes for multiplication factors and reaction rate ratios. Both the homogeneous and heterogeneous pin-cell calculations were conducted for 15 cases. The results of MICROX-2 calculations show good agreement with the reference values. The arithmetic average errors of k∞ for the homogeneous and heterogeneous lattices are 0.30% and 0.44%, respectively. For the finite lattices (five cases for water reactor fuels), the average error of keff is 0.32%. These errors are due to the combined effect of the solution method and the cross-section library. Especially for the fast reactor cases, the prediction of the physics parameter by MICROX-2 deteriorates when the fuel volume increases, which is mostly due to the simplified resonance self-shielding model.

Using TRIGA Mark II research reactor for irradiation with thermal neutrons

Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 1015neutrons/cm2 in irradiation time of 20h.

Accurate Fission Data for Nuclear Safety

The Accurate fission data for nuclear safety (AlFONS) project aims at high precision measurements of fission yields, using the renewed IGISOL mass separator facility in combination with a new high current light ion cyclotron at the University of Jyväskylä. The 30 MeV proton beam will be used to create fast and thermal neutron spectra for the study of neutron induced fission yields. Thanks to a series of mass separating elements, culminating with the JYFLTRAP Penning trap, it is possible to achieve a mass resolving power in the order of a few hundred thousands. In this paper we present the experimental setup and the design of a neutron converter target for IGISOL. The goal is to have a flexible design. For studies of exotic nuclei far from stability a high neutron flux (1012 neutrons/s) at energies 1 – 30 MeV is desired while for reactor applications neutron spectra that resembles those of thermal and fast nuclear reactors are preferred. It is also desirable to be able to produce (semi-)monoenergetic neutrons for benchmarking and to study the energy dependence of fission yields. The scientific program is extensive and is planed to start in 2013 with a measurement of isomeric yield ratios of proton induced fission in uranium. This will be followed by studies of independent yields of thermal and fast neutron induced fission of various actinides.

Improvements and Extensions of the Neutron Cross Section and Fluence Standards

Improvements have been made to the nuclear data standards largely as a result of an IAEA Data Development Project. The work includes the traditional activities related to standards, extending the energy ranges of some standards, and reference data that are not as well known as the standards but can be very useful in the measurements of certain types of cross sections. Also included is an effort to improve evaluations of 235U thermal and 252Cf spontaneous fission neutron spectra.

The molten salt reactor (MSR) in generation IV: Overview and perspectives

Molten Salt Reactors (MSR) with the fuel dissolved in the liquid salt and fluoride-salt-cooled High-temperature Reactors (FHR) have many research themes in common. This paper gives an overview of the international R&D efforts on these reactor types carried out in the framework of Generation-IV. Many countries worldwide contribute to this reactor technology, among which the European Union, France, Japan, Russia and the USA, and for the past few years China and India have also contributed. In general, the international R&D focuses on three main lines of research. The USA focuses on the FHR, which will be a nearer-term application of liquid salt as a reactor coolant, while China also focuses on solid fuel reactors as a precursor to molten salt reactors with liquid fuel and a thermal neutron spectrum. The EU, France and Russia are focusing on the development of a fast spectrum molten salt reactor capable of either breeding or transmutation of actinides from spent nuclear fuel.Future research topics focus on liquid salt technology and materials behavior, the fuel and fuel cycle chemistry and modeling, and the numerical simulation and safety design aspects of the reactor. MSR development attracts more and more attention every year, because it is generally considered as most sustainable of the six Generation-IV designs with intrinsic safety features. Continuing joint efforts are needed to advance common molten salt reactor technologies.

Cold neutron moderator on an upgraded IBR-2 reactor: The first set of results

The first criticality of a new KZ-202 neutron moderator on the IBR-2M reactor is achieved. The moderator consists of thermal and cold units. The former is a room-temperature comb water moderator; the latter, a moderator using a mixture of aromatic hydrocarbons (mesitylene and m-xylene). The cold moderator is filled with granules of this mixture, which are supplied by a cold helium flow, and operates at 30 K. The combination of two units in one moderator makes it possible to simultaneously take the thermal and cold neutron spectra for extracted-beam spectrometers. The arrangement of the thermal and cold moderators is numerically optimized by the Monte Carlo method. The use of the cold moderator allows a 13-fold increase in the cold neutron intensity from its surface.

Comparison of thermal neutron detection efficiency of 6Li scintillation glass and 3He gas proportional tube

We report on a comparison study of the 3He gas proportional tube and the 6Li incorporated scintillation glasses on thermal neutron detection efficiency. Both 3He and 6Li are used commonly for thermal neutron detection because of their high neutron capture absorption coefficient. By using a neutron source 252Cf and a paraffin moderator in an alignment system, we can get a small beam of thermal neutrons. A flash ADC is used to measure the thermal neutron spectrum of each detector, and the detected number of events is determined from the spectrum, then we can calculate the detection efficiency of different detectors. Meanwhile, the experiment has been modeled with GEANT4 to validate the results against the Monte Carlo simulation.

High enriched to low enriched fuel conversion in YALINA Booster facility

The YALINA Booster zero power facility is a subcritical assembly located in Minsk (Belarus). This assembly has the special feature of operating on fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in polyethylene matrix. This study discusses the high enriched to low enriched fuel conversion. In order to increase the neutron multiplication of the assembly loaded with low (21%) enriched fuel in the fast zone, the number of fuel rods in the thermal zone cannot be augmented. Consequently, the effective multiplication factor of the configuration with 21% enriched uranium fuel in the fast zone has been enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located in-between the fast and the thermal zones, to form a circular rather than a square arrangement. The MCNP computer simulation results obtained with the circular arrangement of the absorber rods are in good agreement with the experimental data.

Coupled neutronics/thermal hydraulics evaluation for thorium based fuels in thermal spectrum SCWR

Thorium can supplement the current limited reserves of uranium. In current study, analyses are performed for thorium based fuels in thermal neutron spectrum Super Critical Water Reactor (SCWR). Thorium based fuels are studied in two roles. First role being replacement of conventional uranium dioxide fuel while the other being burner of Reactor Grade Plutonium (RG-Pu) in thermal neutron spectrum SCWR. Coupled neutron physics/thermal hydraulics analyses are performed due to large density variation of coolant over the active fuel length. Analyses reveal that thorium-uranium MOX fuels lead to smaller burnup values as compared to equivalent enriched uranium dioxide but possess the advantage of smaller excess reactivity at Beginning of Life (BOL). This can lead to savings in the form of Burnable Poisons (BP). Smaller fuel average temperature values are obtained for thorium-uranium MOX fuels as compared to uranium dioxide fuel option. Coated fuel option utilizing mixed thorium-uranium mono nitride fuel can help further decrease fuel average temperature values for thorium based fuels. U-233, produced in thorium uranium fuels, contribution towards fission energy produced is smaller as compared to plutonium produced in conventional uranium dioxide fuel. In terms of proliferation resistance, approximately 40% less quantity of plutonium is produced for thorium-uranium MOX fuels (for studied compositions) as compared to equivalent enriched uranium dioxide fuel. But, there is not much difference between the discharged plutonium vector compositions. Thorium–Plutonium based fuels lead to significantly harder spectrum which results in larger spread in radial power density and eventually causes larger values for thermal hydraulic parameters like fuel and clad temperature. Due to almost no production of plutonium, thorium based fuels can be a very good option to burn RG-Pu in thermal spectrum SCWR. Thorium based fuels destroyed almost 74% initially loaded RG-Pu as compared to 60% for uranium based MOX. HEU based thorium fuels can be a very good option for replacing conventional uranium dioxide fuels as very small quantities of plutonium is produced. This option, although, has regulatory issues due to use of HEU material.

Fluctuations of Mass Yield Distribution for the Epithermal Fission of 235U and Its Impact on HCLWR Calculations

The need for accurate nuclear data represents a permanent challenge to improve the calculation tools used in reactor physics. Uranium-235 is one of the most important isotopes, and its related nuclear data need to be known with a high degree of accuracy. In this context, many studies have been undertaken to improve the fission yields of 235U for the main fission products in spent fuels.For a few years, an increasing interest has been observed for high-conversion light water reactors, for better use of the fuel and for nonproliferation considerations. These concepts are based on a low moderation ratio (˜0.9) and the use of highly enriched mixed oxide fuels (>8%). Because of a neutron flux much harder than in pressurized water reactors, calculations require a good knowledge of nuclear data in the epithermal range to accurately predict the fuel depletion with burnup. In particular, the energy dependence of the fission yields must be considered, because of the existence of fluctuations of the fragment fission yields in the resonances of 235U(n,f). Unfortunately, the fluctuations are incorrectly taken into account in calculation codes because of the rough energy description in the usual nuclear data libraries.In this paper, integral experiments are presented for the identification and the quantification of this effect in an almost 1/E neutron spectrum. The experiments consist of the irradiation of 235U samples in the MINERVE reactor (CEA Cadarache), first in a mostly thermal neutron spectrum and second in a mostly epithermal one. Measurements of some abundant fission products are realized with gamma-ray spectrometry and show fluctuations of the 235U cumulative fission yields between the two experiments. Results are interpreted with the multimodal random neck rupture model as a variation of 5% of the weight ratio w1/w2 between the standard I and standard II fission modes, and compared with the differential experiments.

A boron-coated ionization chamber for ultra-cold neutron detection

The design and performance of a boron-coated ionization chamber for the detection of ultra-cold neutrons (UCN) are presented. We detect UCN from the solid deuterium-based UCN source at the Los Alamos Neutron Science Center. Our results indicate comparable efficiency to 3He ionization chambers and proportional counters currently used at the UCN source. In addition, the ion chamber is used to detect thermal neutrons; a comparison of the thermal neutron and UCN pulse-height spectra indicates that UCN mostly capture near the layer surface.

Updating and Extending the IRDF-2002 Dosimetry Library

Abstract The International Reactor Dosimetry File (IRDF)-2002 released in 2004 by the IAEA (see http://www-nds.iaea.org/irdf2002/) contains cross-section data and corresponding uncertainties for 66 dosimetry reactions. New cross-section evaluations have become available recently that re-define some of these dosimetry reactions including: (1) high-fidelity evaluation work undertaken by one of the authors (KIZ); (2) evaluations from the US ENDF/B-VII.0 and candidate evaluations from the US ENDF/B-VII.1 libraries that cover reactions within the International Evaluation of Neutron Cross-Section Standards; (3) European JEFF3.1 library; and (4) Japanese JENDL-4.0 library. Additional high-threshold reactions not included in IRDF-2002 (e.g.,59Co(n,3n) and 209Bi(n,3n)) have been also evaluated to characterize higher-energy neutron fields. Overall, 37 new evaluations of dosimetry reactions have been assessed and intercomparisons made with integral measurements in reference neutron fields to determine whether they should be adopted to update and improve IRDF-2002. Benchmark calculations performed for newly evaluated reactions using the ENDF/B-VII.0 235U thermal fission and 252Cf spontaneous fission neutron spectra show that calculated integral cross sections exhibit improved agreement with evaluated experimental data when compared with the equivalent data from the IRDF-2002 library. Data inconsistencies or deficiencies of new evaluations have been identified for 63Cu(n,2n), 60Ni(n,p)60m+g Co, 55Mn(n,γ), and 232Th(n,f) reactions. Compared with IRDF-2002, the upper neutron energy boundary was formally increased from the actual maximum energy of typically 20 MeV up to 60 MeV by using the TENDL-2010 cross sections and covariance matrices. This extension would allow the updated IRDF library to be also used in fusion dosimetry applications. Uncertainties in the cross sections for all new evaluations are given in the form of relative covariance matrices. Newly evaluated excitation functions should be considered as suitable candidates in the preparation of an improved version of the IRDF that was planned to be released for testing in December 2011.

Monte Carlo simulations for the analysis of texture and strain measured with Bragg edge neutron transmission

The transmission spectrum of cold and thermal neutrons displays sudden intensity changes, known as Bragg edges, whose appearance is connected to the crystallographic structure of the investigated sample. However, many difficulties in interpreting and understanding the detected signal exist, in particular when results are evaluated quantitatively. We measured strain and texture properties of different samples utilizing the Bragg edge method on the CONRAD and VSANS instruments at Helmholtz Centre Berlin as well as on ENGIN-X at ISIS, UK. The different results will be compared and validated with conventional diffraction measurements and theoretical considerations. According to this, a Bragg edge neutron transmission simulation was developed to include instrument effects together with sample characteristics. Therefore, a neutron ray-tracing simulation package was applied to emulate the employed instruments in combination with the Bragg edge transmission technique. A new sample module has been created to realize energy-dependent transmission simulations based on neutron cross sections calculations. Towards the analysis of the conducted experiments the virtual sample features the expected strain and texture parameters. The results of the simulation are in good agreement with the experiments and will be discussed with regard to the prospects and limitations of this measurement method and the utilized instruments.

MAX phase carbides and nitrides: Properties for future nuclear power plant in-core applications and neutron transmutation analysis

Abstract A family of ternary carbides and nitrides, known as MAX phases, combine attractive properties of both ceramics and metals, and has been suggested for potential nuclear reactor applications. The unirradiated materials properties of importance for in-core structural materials and as fuel pellet coatings for several leading MAX phase materials have been summarized from literature. The materials show high mechanical damage tolerance in terms of creep, thermal/mechanical fatigue and fracture resistance, and very good chemical compatibility with select coolants such as molten lead and sodium. Neutron activation has been calculated for commercial purity materials exposed to both idealized fast and thermal reactor neutron spectra for 10, 30, and 60 years of exposure. The specific activities of Ti 3 SiC 2 , Ti 3 AlC 2 , and Ti 2 AlC were compared to those of SiC and Alloy 617, two leading candidate materials for next generation reactor components. The specific activities of MAX phases were similar to SiC and three orders of magnitude less than Alloy 617 after 10–60 years decay for all three activation times in both the fast and thermal spectra. As with SiC, the main radioisotopes after a decay period of 10 years for all three activation times in the MAX phases are tritium and C 14 . Neutron irradiation results of Ti 3 SiC 2 , Ti 3 AlC 2 , and Ti 2 AlC experimentally confirmed the neutron transmutation analysis.

Development of Pyrochemical Reprocessing of the Spent Nuclear Fuel and Prospects of Closed Fuel Cycle

Molten-Salt Reactor (MSR) is a design of an advanced reactor system from the GEN IV family working in thermal or epithermal neutron spectrum and using thorium or transuranium fuel in the form of molten fluorides. It is based on the experience with the development of the molten-salt reactor technology in the Oak-Ridge National Laboratory in the United States. The MSR fuel cycle with integrated reprocessing represents one of the potential ways both for significant decrease of total amount of radioactive wastes for final deposition and for utilization of nuclear energy for electricity and heat production as effectively as possible. There are two pyrochemical reprocessing techniques studied in NRI Rez plc which are considered to be applied both for reprocessing of already existing spent fuel and for preparation and „on-line“ reprocessing of MSR fuel: (i) the Fluoride Volatility Method (FVM), which performs chemical conversion of spent thermal oxide fuel components into fluorides and their consequent separation by means of their different volatility, thermal stability and chemical affinity to various sorbents; and (ii) electrochemical separation of the actinides (Ans) and fission products (FP), represented mainly by lanthanides (Lns), from each other by electrolytic deposition method on solid cathode in molten fluoride media.

Benchmark analysis for accelerator-driven subcritical reactor with pulsed neutron generator in Kyoto University Critical Assembly (KUCA)

This study aims to compare the theoretical and experimental results of an accelerator driven system with a sub-critical core. A benchmark analysis of number of experimentations at Kyoto University Critical Assembly (KUCA) has been made. There were two series of experimentations and each series has four cases. The main objective of these experiments was to study the sub-critical neutronics characteristics (neutron spectrum, reaction rates distributions) and validate measurements techniques appropriate to sub-critical systems. The KUCA core is a thermal neutron spectrum assembly, fuelled with highly enriched uranium fuel, and moderated and reflected by polyethylene. The neutron generator consisted of a Cockcroft-Walton type machine accelerating deuterons onto a tritium target yielding 10 9 n/s. Theoretical calculations are done with MCNP-5 1.4v with cross-section library ENDF-B-VI-5. Subcriticality, excess reactivity for different core configurations were measured and compared with theoretical calculations. Reaction rates of indium wire from source to inside of the core were compared experimentally and theoretically. In general a good comparison has been seen in this work.

Comparison of MCB and FISPACT burn-up performances using the HELIOS experiment technical specifications

Abstract The objective of the HELIOS experiment was to test innovative, uranium-free nuclear fuel containing americium and to study helium release and fuel swelling behaviour during irradiation. The experiment consisted of irradiation of five fuel samples, each containing various initial concentrations of Pu and Am in different inert matrices. The irradiation was performed in the High Flux Reactor (HFR) at Petten during nine cycles (i.e. about nine months) for a total duration of 240 equivalent full-power days. The neutronic calculations play an important role in preparation and analysis of the irradiation experiments. Therefore, this paper presents an explicit burn-up analysis of the HELIOS samples using two numerical tools, the MCB and FISPACT codes. The focus of this calculation is on the concentrations of Pu, minor actinides (MAs) and He during irradiation time and on total sample power due to nuclear reactions and decay. Particular attention was paid to Am242m, the fission cross-section of which is significant in the thermal neutron spectrum. The MCB code incorporates an improved method of producing this, which was subsequently tested and qualified. The sensitivity of the nuclide evolution to the choice of JEF2.2 and JEFF3.1 transport cross-section libraries was also investigated.

Statistical Model Analysis of (n,p) Cross Sections Averaged over the Fission Neutron Spectrum

Experimental (n, p) cross sections averaged over the thermal fission neutron spectrum of U-235 are analyzed in the framework of the statistical model and some regularities are observed. The experimental data are described by the statistical model in which the evaporation model and the constant nuclear temperature approximation are used. Also, the effective average neutron energy for fission spectrum of U-235 in the case of (n, p) cross sections was found to be around 4 MeV.

Monte Carlo simulation of neutron tomography for palm weevil detection

The use of neutron for Non Destructive Imaging (NDI) techniques has many advantages over other (NDI) methods. Using well-established X-ray imaging techniques can provide easy and direct results with some limitations where the sensitivity for light elements is very low. On the other hand, neutron is highly sensitive to water content and can provide extra qualitative information. Comparing the results of the two imaging techniques are investigated in this work with the aim of identifying the palm weevil. At larva stage of the weevil’s life it is characterized by highly water content in the trunk of the palm tree which itself composed of spongy watery texture in some types of palm tree. MCNPX 2.5.0 code with neutron radiography tally was used to obtain the 2D projection then reconstructed to 3D tomography image using OSCaR post processing package. The neutron and photon mesh tallies is utilized to study the neutron and photon fluences from monoenergetic thermal neutron beam and neutron spectrum. There are fundamental difficulties in neutron detection which result in misleading information arises from neutron scattering when constructing cone beam CT neutron images, however, neutron radiography provide better methods for the weevil detection from 2D projection.

Experiments with ultracold neutrons

Ultracold neutrons (UCN) form a tiny low-energy fraction of the Maxwellian spectrum of thermal neutrons in moderators of nuclear reactors and spallation sources. Their energy is extremely low (∼10−7 eV), their velocity is a few meters per second, and their effective temperature is as low as ∼1 mK. The specific feature of UCN is their nearly total elastic reflection by the nuclear-optical potential of many materials at any incidence angle, so that they could be stored in closed traps for many minutes, and used for extremely sensitive measurements. The fraction of UCN in a thermal neutron flux is as low as 10−12–10−11, and serious efforts are under way all over the world to produce UCN in larger amounts. UCN are widely used in precision particle physics experiments. Applications of UCN are emerging in surface and nanoparticle physics. Here we focus on recent advances in the field.