Neutron Absorber Materials Research Articles

Synthesis and evaluation of thermal neutron attenuation properties of lithium orthosilicate for its application as a beam shaping material on BNCT facilities

Thermal neutron attenuation capacity of Li4SiO4 was evaluated to assess its potential capabilities as a beam shaping material for boron neutron capture therapy (BNCT) facilities. Samples of Li4SiO4 were prepared by two different synthesis methods, using different raw materials and were characterized using x-ray, electron diffraction and transmission electron microscopy. Neutron measurements were performed at the BNCT and the neutron radiography facilities of Centro Atómico Bariloche. Considering its natural isotopic abundance, Li4SiO4 proved to be remarkably effective in comparison with other neutron-absorbing materials. Given the availability of natural Lithium in local salt mines and the scalable feasibility, Li4SiO4 qualifies as a potential material for BNCT beam shaping applications.

Microstructure and mechanical properties of (B4C+Al2O3)/Al composites designed for neutron absorbing materials with both structural and functional usages

To meet the demand for the new generation of neutron absorber materials (NAMs) for the dry storage of the spent nuclear fuels, (B4C + Al2O3)/Al composites were fabricated by powder metallurgy technique using ultrafine Al powders. The composites designed with various fabricating parameters and fabricated at various sintering temperatures were characterized by electron microscopy and mechanically tested. The sample sintered at 450 °C shows the best strength-ductility balance at 350 °C (106.2 MPa in ultimate tensile strength and 9.6% in elongation). Addition of B4C particles and increase of the Al2O3 film thickness could enhance the strength of the composites at room temperature but showed no obvious effect on the strength at 350 °C. When sintering temperature of the composites increased from 450 °C to 550 °C, the transformation of amorphous Al2O3 lamellae to γ-Al2O3 particles led to deterioration of the strength of the composites. Based on the analyses of both high-temperature deformation mechanism and strengthening mechanism, it was considered that the amorphous Al2O3 could pin the grain boundaries and prevent them from gliding, which was the main factor to significantly increase the high-temperature strength. Based on the results, a strategy to design the aluminium matrix NAMs with excellent high-temperature strength was proposed.

Enhancing high-temperature strength of B4C–6061Al neutron absorber material by in-situ Mg(Al)B2

The industrial development of dry storage method of spent nuclear fuels strongly requires the low-cost B4C–Al neutron absorber materials (NAMs) with high strength and high stability at elevated temperatures. In the present work, the B4C–6061Al composites were fabricated by powder metallurgy method with different hot-pressing temperatures. Microscopic analysis showed that as hot-pressing temperatures increased from 560 °C to 630 °C, the interfacial reactions became severe, resulting in the formation of Al3BC and Mg(Al)B2 compounds. The mechanical properties of the composites were enhanced by the interfacial reactions, specifically, the tensile strength of the sample hot-pressed at 630 °C was 40–59% higher than that at 560 °C at testing temperatures from room temperature up to 350 °C. Moreover, the high-temperature-pressed sample maintained the invariant strength level after suffering annealing at 400 °C for 8000 h. Characterization of the fractured composite indicated that the Mg(Al)B2 nanorods which effectively impeded the dislocation movement and grain boundary sliding provided a significant contribution to the composite strength.

The effect of boron and titanium addition on the behavior of steel alloys of base composition AISI304 as a nuclear radiation shielding material

Borated stainless steel alloys are candidate neutron absorber materials for criticality control in disposal containers or dry storage canisters. Type 304 borated stainless steels are similar to conventional Type 304 stainless steels except that they contain a boron addition, which imparts a much higher thermal neutron absorption cross section than other austenitic stainless steels. Three grades of steel alloys with base composition of AISI304 (SS304Ti, SS304B and SS304TiB) were modified and produced using 30 kg pilot plant medium frequency induction furnace at the same conditions. The macroscopic-cross sections for neutrons higher than 10 keV, slow, and total slow neutrons were measured using 241Am-Be neutron source for the produced samples. The developed samples were found to have higher cross sections for neutrons higher than 10 keV, slow, and total slow neutrons than SS304. Additionally, gamma ray shielding properties were measured for several gamma ray energies that emitted from 232Th radioactive source. Among the investigated steels, the lowest corrosion rate was found in the modified steel containing B. Samples of the properly treated steels were subjected to microstructure observation, hardness, tensile and impact testing. The microstructure observation revealed an austenitic phase in all investigated steel alloys and a very large number of twins in the optical micrographs of the investigated samples. The strengthening effect of B or Ti and B is to great extent the reason of the solution hardening and precipitation strengthening effects of these elements as these elements are strong carbide or nitride formers.

Qualification Requirements Design for Neutron Absorber Plate for Spent Fuel Racks in Domestic Nuclear Power Plants

Boron carbide (B4C) particle-reinforced aluminum matrix composite is the key material for use as neutron absorber plate in fuel storage applications for Generation III advanced passive nuclear power plants in China. This material has once depended upon importing with various restrictions so that it has meaningful practical significance to realize the localized manufacturing for this material in China. More importantly, since it is the first time for this material to be used in domestic plant, particular care should be taken to assure the formal supplied products exhibit high stabilized and reliable service in domestic nuclear engineering. This paper initiates and proposes a principle design framework from technical view in qualification requirements for this material so as to guide the practical engineering application. Aiming at neutron absorber materials supplied under practical manufacturing condition in engineering delivery, the qualification requirements define B4C content, matrix chemistry, 10B isotope, bulk density, 10B areal density, mechanical property, and microstructure as key criteria for material performance. The uniformity assessment as to different locations of this material is also required from at least three lots of material. Only qualified material meeting all of the qualification requirements should proceed to be verified by lifetime testing such as irradiation, corrosion, and thermal aging testing. Systematic and comprehensive performance assessments and verification for process stabilization could be achieved through the above qualification. The long-term service for this neutron absorber material in reliable and safe way could be convincingly expected in spent fuel storage application in China.

Corrosion onset associated with the reinforcement and secondary phases in B4C-6061Al neutron absorber material in H3BO3 solution

The microstructures in B4C-6061Al composite hot-pressed at different temperatures (560 °C, 620 °C) are characterized, and the corrosion behavior of the materials is studied using electrochemical tests and quasi in-situ TEM observations. The composite fabricated at 620 °C is composed of more matrix-reinforcement reaction products (Al3BC at the interfaces and MgB2 in the alloy matrix). Corrosion occurred slightly in boric acid but severely in Cl−-containing solution. The B4C/Al interface in both samples is preferentially attacked and the MgB2 dispersions in the high-temperature-pressed composite remarkably promoted the pitting. Dealloying of Mg2Si precipitate does not obviously deteriorate the corrosion resistance of the composite.

Development of integral type spent fuel pool storage rack with gadolinium and europium-containing structure materials

This study presents the development of a new integral-type rack design, characterized by the use of gadolinium (Gd)-containing structure materials that enhances the capacity of spent fuel (SF) pool storage by exploiting the high neutron absorption capability of Gd. Appropriate types and contents of Gd-based neutron-absorbing materials are selected for the new design through parametric studies. For high-reactivity fuels (region I of SF storage pools), neutron-absorbing material composed of Gd 0.7 at% with Eu 2.73 at% is found to be an optimal neutron absorber whereas for low-reactivity fuels (region II of SF storage pools), a composition of Gd 0.7 at% with Eu 8.38 at% is found to be optimal. A criticality safety analysis shows that the newly designed racks are more subcritical than conventional racks for both regions I and II. The additional reactivity margin yielded by the new integral-type design can be used to reduce the pitch of the rack while maintaining equivalent subcriticality compared to conventional rack design. This study demonstrates the potential of Gd-based neutron absorbers in structure materials for increasing the total amount of fuel assemblies that can be stored in a SF storage pool.

Influence of boron enrichment in control rods on neutron-physical characteristics of the core of the reactor VVER

Boron is widely used in nuclear power as a neutron absorber material, thereby creating the possibility of controlling a nuclear reactor by changing the neutron multiplication factor. Natural boron contains about 20% 10B and 80% 11B with absorption sections respectively 3840 barn and 0.05 barn. The paper deals with changes in neutron-physical parameters of the reactor core of VVER type when the isotope content of 10B in boron carbide of absorber rods (AR) of the control and protection system (CPS) and in boric acid solution in the coolant of the VVER reactor is varied, and also analyzes the practicality of increasing the enrichment of 10B isotope in the control rods (CR). The model of VVER-1200 fuel assembly (FA) was used for calculations. In the course of work were modelled and the calculations of the three models of the coolant with different boric acid concentrations, as well as two models of AS with different enrichment in the isotope 10B.

Control rod treatment for FEM based radiation transport methods

This paper presents a novel control rod treatment for finite element method (FEM) based reactor physics simulations. The new method falls into the class of homogenization methods which treat material interfaces in mesh elements by defining a representative, homogenized nuclear cross section set. In contrast to existing homogenization based cusping treatments that attempt to approximate flux-volume weighting of the cross sections, we propose to exactly integrate terms appearing in the variational form of the transport or diffusion equation over mixed material elements. This goal is achieved by defining smoothly varying cross sections in elements that straddle material boundaries. It is shown that the smoothly varying cross sections can be obtained by projecting the piecewise set of cross sections onto the set of Legendre polynomials up to a finite order. Implementation of smoothly varying cross sections is supported within the Rattlesnake radiation multiphysics code at the Idaho National Laboratory. Homogenization methods traditionally suffer from cusping effects that originate from overestimation of absorption cross sections in strongly neutron-absorbing materials as found in control rods. We demonstrate that the presented method is (1) highly effective in removing cusping effects and (2) provides accurate results based on the C5G7-TD benchmark, a time-dependent version of the C5G7 benchmark.

Evaluation of damage in neutron irradiated boron carbide

When irradiated in a reactor, neutron absorber materials undergo two different damages induced, first, by the elastic scattering of neutrons; second by the neutron absorption reactions. In boron carbide irradiated in a fast neutron flux, neutron scattering and the slowing-down of He and Li arising from the (n,α) absorption reaction both lead to the displacement of B and C atoms which energy ranges up to a few MeV. Simulating the neutron damage by ion irradiation requires the calculation of the damage produced in both cases. In this paper we propose an estimation of the actual defect production rate resulting from the fast neutron irradiation. For this, we first estimate the energy spectra of both the primary knocked-on atoms and atoms created by the absorption reactions, here in a Phenix-like neutron spectrum. We then deduce from SRIM calculations the actual energy distribution of all the atoms displaced along the displacements cascades induced by those primary projectiles. At last, we obtain an estimation of the number of displaced atoms per produced helium, about 305, most of them resulting from neutron scattering. This is far from a Kinchin-Pease or NRT estimation, of the order of several thousands, this arising from the fact that most of the energy is dissipated through electronic interactions. Such results are then used in order to perform ion irradiations aiming at a realistic simulation of synergetic effects of helium implantation and damage production.

A new methodology for studying neutron absorber materials: First results with boron carbide

Abstract This paper presents a new methodology used to study transmutation damage in boron carbide (B4C). B4C samples were set in a specially designed sample holder and irradiated in a neutron thermal flux in the ILL reactor. After describing the sample holder and the irradiation conditions, this paper discusses the results of the first post-irradiation examinations by secondary-ion mass spectrometry (SIMS), X-ray diffraction (XRD) and scanning electron microscopy (SEM). The results obtained for a B4C disc with natural boron composition prove the efficiency of the methodology.

First-principles investigation of neutron-irradiation-induced point defects in B4C, a neutron absorber for sodium-cooled fast nuclear reactors

Boron carbide (B4C) is a leading candidate neutron absorber material for sodium-cooled fast nuclear reactors owing to its excellent neutron-capture capability. The formation and migration energies of the neutron-irradiation-induced defects, including vacancies, neutron-capture reaction products, and knocked-out atoms were studied by density functional theory calculations. The vacancy-type defects tend to migrate to the C–B–C chains of B4C, which indicates that the icosahedral cage structures of B4C have strong resistance to neutron irradiation. We found that lithium and helium atoms had significantly lower migration barriers along the rhombohedral (111) plane of B4C than perpendicular to this plane. This implies that the helium and lithium interstitials tended to follow a two-dimensional diffusion regime in B4C at low temperatures which explains the formation of flat disk like helium bubbles experimentally observed in B4C pellets after neutron irradiation. The knocked-out atoms are considered to be annihilated by the recombination of the close pairs of self-interstitials and vacancies.

Monte Carlo simulation of biological shielding parameters in PGNAA

A prompt γ neutron activation analysis (PGNAA) facility based on a 252 Cf spontaneous fission neutron was performed with the final goal of obtaining various optimised radiation protection design parameters. MCNP Monte Carlo code was used to simulate the lead (Pb) shield thickness on two sides of the sample chamber, and B4 C, B2O3, LiOH, LiF, and borax were selected as neutron-absorbing materials that were simulated to determine optimal parameters for the optimal design of a biological multilayer composite shield. A series of calculations performed with the MCNP code indicated that the 5 cm thick Pb innermost layer of a biological shield rapidly reduced fast neutrons, and a moderation layer of 25 cm thick paraffin, 20 mm thick borax, and 2 mm thick B4 C were selected as neutron-absorbing materials. The calculated biological shield protection parameters satisfied the requirements of the PGNAA facility used in the experiment.

Neutron tomography simulation by MAVRIC/Monaco code

In this study, MAVRIC/Monaco module integrated with SCALE6.1 package is used to elucidate experimental neutron transmission for various neutron-absorbing materials. Neutron beam hardening and its effect on shielding calculation and tomographic image reconstruction accuracy are presented.The experimental data for studied materials are used for comparison purpose and to prove the capability of MAVRIC/Monaco code in neutron tomography simulation. The good agreement between experimental and calculated values confirms the capability of SCALE6.1 package in obtaining physical data exactly analogous to an experimental measurement.

Performance evaluation of METAMIC neutron absorber in spent fuel storage rack

Abstract High-density spent fuel (SF) storage racks have been installed to increase SF pool capacity. In these SF racks, neutron absorber materials were placed between fuel assemblies allowing the storage of fuel assemblies in close proximity to one another. The purpose of the neutron absorber materials is to preclude neutronic coupling between adjacent fuel assemblies and to maintain the fuel in a subcritical storage condition. METAMIC neutron absorber has been used in high-density storage racks. But, neutron absorber materials can be subject to severe conditions including long-term exposure to gamma radiation and neutron radiation. Recently, some of them have experienced degradation, such as white spots on the surface. Under these conditions, the material must continue to serve its intended function of absorbing neutrons. For the first time in Korea, this article uses a neutron attenuation test to examine the performance of METAMIC surveillance coupons. Also, scanning electron microscope analysis was carried out to verify the white spots that were detected on the surface of METAMIC. In the neutron attenuation test, there was no significant sign of boron loss in most of the METAMIC coupons, but the coupon with white spots had relatively less B-10 content than the others. In the scanning electron microscope analysis, corrosion material was detected in all METAMIC coupons. Especially, it was confirmed that the coupon with white spots contains much more corrosion material than the others.

Hypoeutectic melting in the UO2-x-Gd2O3 system

Gadolinium is one of the best neutron absorber materials and its use can be considered as a sacrificial material in a Sodium Fast Reactor core catcher in view of preventing recriticallity. A series of experiments have been conducted in the VITI induction-heated facility to study the melting in the UO2-x-Gd2O3 system with 60–87 mol% gadolinia. These experiments have indicated that the eutectic composition is around 92 mol% Gd2O3 – 8 mol% UO2-x and that the liquidus line is close to that of Popov et al. [Atom. Energ. 110 (2011) pp. 221–229] phase diagram.

Neutron measurements in radiotherapy: A method to correct neutron sensitive devices for parasitic photon response

One of the major causes of secondary malignancies after radiotherapy treatments are peripheral doses, known to increase for some newer techniques (such as IMRT or VMAT). For accelerators operating above 10MV, neutrons can represent important contribution to peripheral doses. This neutron contamination can be measured using different passive or active techniques, available in the literature. As far as active (or direct-reading) procedures are concerned, a major issue is represented by their parasitic photon sensitivity, which can significantly affect the measurement when the point of test is located near to the field-edge. This work proposes a simple method to estimate the unwanted photon contribution to these neutrons. As a relevant case study, the use of a recently neutron sensor for “in-phantom” measurements in high-energy machines was considered. The method, called “Dual Energy Photon Subtraction” (DEPS), requires pairs of measurements performed for the same treatment, in low-energy (6MV) and high energy (e.g. 15MV) fields. It assumes that the peripheral photon dose (PPD) at a fixed point in a phantom, normalized to the unit photon dose at the isocenter, does not depend on the treatment energy. Measurements with ionization chamber and Monte Carlo simulations were used to evaluate the validity of this hypothesis. DEPS method was compared to already published correction methods, such as the use of neutron absorber materials. In addition to its simplicity, an advantage of DEPs procedure is that it can be applied to any radiotherapy machine.

Five-Axis Ultrasonic Additive Manufacturing for Nuclear Component Manufacture

Ultrasonic additive manufacturing (UAM) is a three-dimensional metal printing technology which uses high-frequency vibrations to scrub and weld together both similar and dissimilar metal foils. There is no melting in the process and no special atmosphere requirements are needed. Consequently, dissimilar metals can be joined with little to no intermetallic compound formation, and large components can be manufactured. These attributes have the potential to transform manufacturing of nuclear reactor core components such as control elements for the High Flux Isotope Reactor at Oak Ridge National Laboratory. These components are hybrid structures consisting of an outer cladding layer in contact with the coolant with neutron-absorbing materials inside, such as neutron poisons for reactor control purposes. UAM systems are built into a computer numerical control (CNC) framework to utilize intermittent subtractive processes. These subtractive processes are used to introduce internal features as the component is being built and for net shaping. The CNC framework is also used for controlling the motion of the welding operation. It is demonstrated here that curved components with embedded features can be produced using a five-axis code for the welder for the first time.

Application of a Dynamic-Nanoindentation Method to Analyze the Local Structure of an Fe-18 at.% Gd Cast Alloy

Abstract A dynamic nanoindentation method was applied to study an Fe-18 at.% Gd alloy as a neutron-absorbing material prepared by vacuum arc-melting and cast in a mold. The Fe-18 at.% Gd cast alloy had a microstructure with matrix phases and an Fe-rich primary dendrite of Fe9Gd. Rietveld refinement of the X-ray spectra showed that the Fe-18 at.% Gd cast alloy consisted of 35.84 at.% Fe3Gd, 6.58 at.% Fe5Gd, 16.22 at.% Fe9Gd, 1.87 at.% Fe2Gd, and 39.49 at.% β-Fe17Gd2. The average nanohardness of the primary dendrite phase and the matrix phases were 8.7 GPa and 9.3 GPa, respectively. The fatigue limit of the matrix phase was approximately 37% higher than that of the primary dendrite phase. The dynamic nanoindentation method is useful for identifying local phases and for analyzing local mechanical properties.

Heat transfer characteristics and operation limit of pressurized hybrid heat pipe for small modular reactors

In this paper, a hybrid heat pipe is proposed for use in advanced nuclear power plants as a passive heat transfer device. The hybrid heat pipe combines the functions of a heat pipe and a control rod to simultaneously remove the decay heat generated from the core and shutdown the reactor under accident conditions. Thus, the hybrid heat pipe contains a neutron absorber in the evaporator section, which corresponds to the core of the reactor pressure vessel. The presence of the neutron absorber material leads to differences in the heated diameter and hydraulic diameter of the heat pipe. The cross-sectional areas of the vapor paths through the evaporator, adiabatic, and condenser sections are also different. The hybrid heat pipe must operate in a high-temperature, high-pressure environment to remove the decay heat. In other words, the operating pressure must be higher than those of the commercially available thermosyphons. Hence, the thermal performances, including operation limit of the hybrid heat pipe, were experimentally studied in the operating pressure range of 0.2–20bar. The operating pressure of the hybrid heat pipe was controlled by charging the non-condensable gas which is unused method to achieve the high saturation pressure in conventional thermosyphons. The effect of operating pressure on evaporation heat transfer was negligible, while condensation heat transfer was affected by the amount of non-condensable gas in the test section. The operation limit of the hybrid heat pipe increased with the operating pressure. Maximum heat removal capacity of the hybrid heat pipe was up to 6kW which is meaningful value as a passive decay heat removal device in the nuclear power plants. Based on the experimentally measured maximum heat removal capacities, models predicting the operation limit (flooding limit) of the hybrid heat pipe were developed.