Beryllium Blocks Research Articles

Analysis of the excess reactivity and control rod worth of RSG-GAS equilibrium silicide core using Continuous-Energy Monte Carlo Serpent2 code

This present work is a continuation of our previous work to verify the RSG-GAS equilibrium silicide core parameters. Earlier, the burnup of 22 fuel elements in the 88th equilibrium core has been measured, and it showed a good agreement between the experimental and calculation results. In this paper, several important neutronics parameters documented from the experiment at the same core, such as the excess reactivity, integral control rod worth, and total control rod worth, are confirmed using the full core calculation results of Monte Carlo Serpent2 code in conjunction with ENDF/B-VII.1 and the most recent ENDF/B-VIII.0 nuclear data libraries. The excess reactivity and integral control rod worth were measured by the positive and negative reactivity compensation method. It is worth mentioning that the measured integral control rod worth hasn’t been verified because the in-core fuel management code of RSG-GAS (BATAN-FUEL) solves only for a 2-dimensional configuration. The fuel element burnup at the end of cycle calculated by BATAN-FUEL and Serpent2 is also compared and discussed. The 3-D Serpent2 model of RSG-GAS first core is modified, and several improvements in the models are proposed, including the explicit modeling of 6 neutron beam tubes in the beryllium block reflector. It is found that the calculated integral control rod worth has a good agreement with the experiment, while Serpent2 overestimates the excess reactivity by a maximum of 661 pcm. The impact of nuclear data is insignificant for the selected core parameters.

Calculation of Design-Basis Accidents Loss of Coolant Circulation for the IRT MIFI Reactor

The results of calculations of emergency situations for the IRT MIFI [Moscow Engineering Physics Institute] reactor with high-enrichment uranium fuel and with conversion to IRT-3M fuel assemblies with fuel cores based on the alloy U–9%Mo in an aluminum matrix with 19.7% enrichment are presented. The transient processes due to disruption of coolant circulation during the execution of safety functions are examined. The PARET 7.5 software is used to calculate the transient processes. Special attention is focused on taking account of emergency protection engagement algorithms and the characteristic features of the devices monitoring the thermohydraulic parameters of the reactor in setting the initial data for the calculation of transient processes. This article presents a safety analysis of the IRT MIFI reactor with high-enrichment uranium fuel and with conversion to IRT-3M fuel assemblies with fuel kernels based on the alloy U–9%Mo in an aluminum matrix with 19.7% enrichment [1]. Emergency situations, caused by a disruption of coolant circulation, with the safety functions fulfi lled are examined. Accidents caused by unauthorized insertion of positive reactivity were analyzed at the preceding stage of this work. The PARET 7.5 software was used to calculate transient processes occurring during an accident. PARET consists of computer programs for performing combined neutron-physical and thermohydraulic calculations based on point neutron kinetics with feedbacks and one-dimensional hydrodynamics and heat transfer [3]. The program includes a library of thermophysical properties of the coolant (water or heavy water) and relations for the coeffi cient of heat emission in different regimes at low pressure, temperature, and coolant velocity which are characteristic for research reactors. Regimes with forced and natural circulation of the coolant are modeled. In the computational model the core can be represented in the form of several regions (channels) with prescribed power density, mass velocity of the coolant, and thermohydraulic parameters. A channel comprises one fuel element (one-dimensional plate or cylinder) with the surrounding coolant. A transient process is modeled by setting the time-varying, externally introduced, reactivity or the mass velocity of the coolant. A fresh load (10 six- and two eight-tube fuel assemblies) and a working load (10 six- and six eight-tube fuel assemblies) with 235 U burnup averaged over the core 24 and 27% for low- and high-enrichment fuel, respectively, are examined. Initial Data. The IRT MIFI reactor is a swimming-pool, water cooled and moderated thermal reactor with a stationary neutron fl ux. The core consists of 16 IRT-3M fuel assemblies with 90% enrichment in the refl ector, comprised of beryllium and aluminum blocks, and is arranged in a 7.8 m deep pool fi lled with chemically purifi ed water (Fig. 1). The fuel assemblies and blocks of the refl ector are arranged in the top containment of the core. The core support lattice is placed in the bottom containment, which is also used for forming the coolant fl ow through the core. In the plan, the containment at the bottom, just as at the top, comprises a rectangular shell. A cylindrical connecting side piece with a fl ange connects the bottom containment with an ejector. The fi rst-loop pressure and intake pipelines enter the containment at the top of the pool. Inside the pool the pressure pipeline is brought into the ejector. The intake pipeline terminates 4 m above the bottom of the pool.

Anisotropic swelling behavior of hot-extruded beryllium

The lifetime of beryllium reflector assemblies is usually determined by neutron irradiation induced swelling, which results in mechanical interferences or fractures of the beryllium elements. Therefore, the dimensional stability and microstructure variations of beryllium during irradiation are important issues to study. In this paper, the microstructure characteristics of S-200-F and EHP-56 beryllium blocks, which were manufactured by using vacuum hot pressing (VHP) and hot extrusion (HE), respectively, were investigated. BeO distributions, grain shapes, and preferred orientations were investigated by using SEM-EPMA and SEM-EBSD systems. Dissimilarly to S-200-F, a strong fiber texture developed in the EHP-56 during the HE process; the basal planes in the majority of grains were arranged along the extrusion direction. To emulate the microstructure evolution during neutron irradiation, we irradiated the electro-polished surface of EHP-56 with protons at room temperature, where the acceleration voltage and the number of protons were 120 keV and 2.0 × 1018 ions/cm2, respectively. Irradiation-induced cavities were observed to be considerably longer along the basal plane in the EHP-56 specimen. Correspondingly, the amount of dimensional change was smaller along the direction parallel to the basal plane.

The Characteristics and Irradiation Capabilities of MARIA Research Reactor in NCBJ Świerk

The high flux research reactor MARIA is a water and beryllium moderated reactor of a pool type with graphite reflector and pressurised channels containing concentric six-tube assemblies of fuel elements. It has been designed to provide high degree of flexibility. The fuel channels are situated in a matrix containing beryllium blocks and enclosed by lateral reflector made of graphite blocks in aluminium cans. The MARIA reactor is equipped with vertical channels for irradiation of target materials, a rabbit system for short irradiations and six horizontal neutron beam channels. The MARIA reactor reached its first criticality in December 1974. The reactor was in operation until 1985 when it was shut down for modernization. The modernization encompassed refurbishment and upgrading of technological systems such as: enlargement of beryllium matrix, inspection of the graphite bocks, upgrading of ventilation and temperature systems.

Effect of sweep gas species on tritium release behavior from lithium titanate packed bed during 14 MeV neutron irradiation

In a fusion reactor, the prediction of tritium release behavior from breeder blanket is important to design the tritium recovery system, but the amount of tritium generated is necessary information to do that. Hence, tritium generation and recovery studies on lithium ceramics packed bed have been started by using fusion neutron source (FNS) in Japan Atomic Energy Agency (JAEA). Lithium titanate (Li2TiO3) was selected as tritium breeding material, and its packed bed was enclosed by the beryllium blocks, and was kept at certain temperature during fusion neutron irradiation. During irradiation, the packed bed was purged with the sweep gas continuously, and tritium released was trapped in each gas absorber selectively by chemical form. In this work, the effect of sweep gas species on tritium release behavior was investigated. In the case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in the case of sweep by helium without water vapor, tritium in gaseous form was released first, and release of tritium in water form was delayed from gaseous tritium and was gradually increased.

Determination of the content of radioactive elements in irradiated beryllium

The elemental composition of the material of beryllium blocks of SM reactor reflector in the initial state was determined by inductively coupled plasma atom emission spectrometry (ICP AES). The α-, β-, and Γ-active components of the radiation from Be irradiated to a neutron fluence of 6 × 1022 cm−2 (E > 0.1 MeV) were determined. Radiochemical analysis revealed the presence in the irradiated Be of 60Co, 54Mn, 137Cs, 134Cs, and 90Sr. The following α-emitters were revealed: 238–241Pu, 241Am,243Am, and 244Cm.

Computational-experimental studies of the neutronics of the SM reactor with different variants of the neutron-trap arrangement

The results of experimental studies of the neutronics of the high-flux SM reactor with different arrangements of the neutron trap are presented. The MCU series of high-precision computer programs implementing the Monte Carlo method is used for computations. Experimental data on reactivity effects, the effectiveness of safety and control rods, and the coefficients of nonuniformity of energy release in the core have been obtained in experiments on a critical assembly ‐ a physical model of the SM reactor ‐ and directly in experiments in the reactor. The error is 4.2‐10% in determining the reactivity parameters and 5‐10% for the relative energy release in the fuel elements. Information on the neutron field formed in the volume of the neutron trap has been obtained for two arrangements of the beryllium and water moderators. The differential and integral energy spectra of the neutrons in the energy interval from 0.5 eV to 20 MeV are obtained for three points inside the trap (external, central series, center). The flux density of thermal, superthemal, and fast neutrons are determined. The SM high-flux research reactor was designed to produce radionuclides and test reactor materials, fuel compositions, and fuel elements [1]. The results of the studies performed with this reactor are used to substantiate the serviceability and safety of nuclear reactors; the radionuclides produced find applications in industry and medicine. The SM reactor has been operated with three different arrangements of the neutron trap. In 1961‐1990, an experimental channel which was connected to an autonomous secondary loop with light water as the coolant was located at the center of the reactor. The next phase of the construction of the reactor, the purpose of which was to increase operational safety, was completed in 1991‐1992 [2]. To increase 252 Cf production, the radionuclide in greatest demand at the time, the water neutron trap was replaced with a centralized beryllium block for holding transuranium targets. The coolant in the primary loop of the reactor started to remove heat from the central block. The cross section of the rebuilt SM reactor is shown in Fig. 1. In 2002, the central beryllium block in the central moderating cavity of the reactor was replaced with a separator structure consisting of 27 14-mm in diameter and 0.5-mm thick zirconium tubes with water in the space between the tubes. The purpose of the change in the construction was to increase production efficiency of 63 Ni, 75 Se, and other isotopes which were in demand at the time and to accumulate which a low-energy neutron spectrum with high flux density is required. The cross section of the central moderating cavity of the SM reactor with different arrangements of the neutron trap is shown in Fig. 2.

Effect of 3He and 6Li accumulation in beryllium blocks on the neutron-physical characteristics of the MIR reactor

Reactions resulting in the accumulation of 3He and 6Li, whose thermal neutron capture cross-section is large, occur under the action of neutron radiation in the beryllium blocks of the MIR reactor core. When a neutron absorber accumulates in the moderator of a reactor, important physical characteristics change: reactivity access, efficiency of safety and control rods, and reactivity effects; in addition, energy release is redistributed. An algorithm for calculating 3H, 3He, and 6Li in each beryllium block of the core has been developed and implemented. This algorithm makes it possible to follow the change in the concentration of these nuclides during reactor operation and shutdown. The 3He and 6Li concentrations are used as initial data for calculating the neutron-physical characteristics of the MIR reactor using the MCU and BERCLI programs. The computational results for the effect of the accumulation of the nuclides indicated on the neutron-physical characteristics of the core are presented.

Use of a Paraffin-Based Grout to Stabilize Buried Beryllium and Other Wastes

The long-term durability of WAXFIX, a paraffin-based grout, was evaluated for in situ grouting of activated beryllium wastes in the subsurface disposal area (SDA), a radioactive landfill at the Radioactive Waste Management Complex, part of the Idaho National Laboratory (INL). The evaluation considered radiological and biological mechanisms that could degrade the grout using data from an extensive literature search and previous tests of in situ grouting at the INL. Conservative radioactive doses for WAXFIX were calculated from the “hottest” (i.e., highest-activity) Advanced Test Reactor beryllium block in the SDA. These results indicate that WAXFIX would not experience extensive radiation damage for many hundreds of years. Calculation of radiation-induced hydrogen generation in WAXFIX indicated that grout physical performance should not be reduced beyond the effects of radiation dose on the molecular structure. Degradation of a paraffin-based grout by microorganisms in the SDA is possible and perhaps likely, but the rate of degradation will be at a slower rate than found in the literature reviewed. The calculations showed the outer 0.46-m (18-in.) layer of each monolith, which represents the minimum expected distance to the beryllium block, was calculated to require 1000 to 3600 yr to be consumed. The existing data and estimations of biodegradation and radiolysis rates for WAXFIX/paraffin do not indicate any immediate problems with the use of WAXFIX for grouting beryllium or other wastes in the SDA.

Performance studies of a mobile neutron source based on the SbBe reaction

A mobile neutron source based on the SbBe reaction for neutron radiography and tomography was designed and set up at the Institut fur Radiochemie, Technische Universitat Munchen, Garching, Germany. The layout is based on both the results of Monte Carlo simulation studies for optimization of the neutron output and the shielding calculations for minimization of the dose rate in the surrounding. The system has dimensions of 58 cm /spl times/ 58 cm /spl times/ 150 cm (w /spl times/ l /spl times/ h) and an overall mass of about 800 kg. It consists of two main components that are combined when operating the source: a lead container for storage and transport of the antimony source (maximum activity 3.7E+11 Bq) and a shielded moderation and collimation unit containing the beryllium block target in which the antimony source is lifted for generating neutrons. Results of experimental investigations of the SbBe source system on the performance and on its applicability for neutron radiography are presented.

Non-destructive analysis of impurities in beryllium, affecting evaluation of the tritium breeding ratio

The non-destructive, pulsed neutron method was used as the most effective way of analyzing the integral effect of impurities in beryllium, relevant to the tritium breeding ratio evaluation. The integral effect was evaluated from time behavior observations of the neutron flux, following the injection of a burst of D–T neutrons into the beryllium assembly. The assembly was constructed from the structural beryllium grade S-200F (Brush Wellman Inc.). Experimental data were compared with the reference data and MCNP-4B calculations. Results show that the measured absorption cross section of thermal neutrons in beryllium blocks is approximately 30% larger than the calculated value, based on the data, specified by the manufacturing company. Impurities in beryllium, such as Li, B, Cd and others, affect the absorption cross section even if the content of impurities is less than 10 ppm.

Isotopic Transmutations in Irradiated Beryllium and Their Implications on Maria Reactor Operation

Beryllium irradiated by neutrons with energies above 0.7 MeV undergoes (n,α) and (n,2n) reactions. The Be(n,α) reaction results in subsequent buildup of 6Li and 3He isotopes with large thermal neutron absorption cross sections causing poisoning of irradiated beryllium. The amount of the poison isotopes depends on the neutron flux level and spectrum. The high-flux MARIA reactor operated in Poland since 1975 consists of a beryllium matrix with fuel channels in cutouts of beryllium blocks. As the experimental determination of 6Li, 3H, and 3He content in the operational reactor is impossible, a systematic computational study of the effect of 3He and 6Li presence in beryllium blocks on MARIA reactor reactivity and power density distribution has been undertaken. The analysis of equations governing the transmutation has been done for neutron flux parameters typical for MARIA beryllium blocks. Study of the mutual influence of reactor operational parameters and the buildup of 6Li, 3H, and 3He in beryllium blocks has shown the necessity of a detailed spatial solution of transmutation equations in the reactor, taking into account the whole history of its operation. Therefore, fuel management calculations using the REBUS code with included chains for Be(n,α)-initiated reactions have been done for the whole reactor lifetime. The calculated poisoning of beryllium blocks has been verified against the critical experiment of 1993. Finally, the current 6Li, 3H, and 3He contents, averaged for each beryllium block, have been calculated. The reactivity drop caused by this poisoning is ~7%.

Neutronics experiments for DEMO blanket at JAERI/FNS

In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li2TiO3 blocks with a 6Li enrichment of 40% and 95%, and beryllium blocks. Sample pellets of Li2TiO3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test specimens simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steel F82H were irradiated as typical fusion materials. The effective cross-sections needed to calculate the formation of the radioactive nuclei (56Co, 184Re, 48V, 206Bi, 65Zn and 51Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections increased remarkably while coming closer to polyethylene board, which was a substitute for water. As a result of this present study, it has become clear that the sequential reaction rates are important factors in the accurate evaluation of induced activity in fusion reactor design.

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.

Thermal neutron flux distribution in ET-RR-2 reactor thermal column

The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a) to provide a thermal neutron flux in the neutron transmutation silicon doping, (b) to provide a thermal flux in the neutron activation analysis position, and (c) to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

Helium and Tritium Behavior in Neutron Irradiated Beryllium

The efficiency of the beryllium application as a plasma-facing material and a neutron multiplier in a solid breeder blanket will depend on helium-induced swelling and tritium and helium release from this metal. The effect of a neutron irradiation on helium and tritium mobility and swelling for three beryllium grades fabricated by VNIINM is described in this paper. The beryllium blocks were irradiated with a neutron fluence (E >0.1 MeV) (2.6 – 3.4). 1021 cm−2 (1.3 – 1.8 dpa) at 550°C, 620°C and 790°C. Mass-spectrometry techniques was used to simultaneously monitoring of gas release during isothermal multi-stage annealing over 500 – 1300°C temperature range.It is shown that the first signs of the helium release have been detected at temperature about 700°C, while the intense tritium release has occurred at all stages of annealing. Based on the data obtained, the diffusion parameters (Do, E) for both the gases in beryllium were calculated. The total amount of helium accumulated in irradiated beryllium varied from 240 appm to 620 appm. The tritium mobility increases significantly when swelling increases, while that for helium changes very slightly. With swelling increase from 0.5 to 1.8 % the ratio of helium to tritium retentions changes approximately from 4:1 to 10:1. The tritium and helium retentions and beryllium swelling are presented as functions of the distance from the irradiated surface. The experimental data are also discussed in comparison with calculations.

Status of EC solid breeder-blanket designs and R&D for DEMO fusion reactors

Within the European Community Fusion Technology Programme, two solid breeder-blankets for a DEMO reactor are being developed. The two blankets have various features in common: helium as coolant and as tritium-purge gas; the martensitic steel MANET as a structural material; and beryllium as the neutron multiplier. The configurations of the two blankets, however, are different: in the breeder inside tube concept, the breeder materials are LiA1O 2 or Li 2 ZrO 3 in the form of annular pellets contained in tubes surrounded by beryllium blocks, with the coolant helium being outside the tubes; in contrast, in the breeder out of tube concept, the breeder and multiplier materials are Li 4 SiO 4 and beryllium pebbles, forming a mixed bed placed outside the tubes containing the coolant helium. The main critical issues for both blankets are the behaviour of the breeder ceramics, and of beryllium under irradiation and the tritium control. Other issues are the low temperature irradiation-induced embrittlement of the MANET; the mechanical effects caused by major plasma disruptions; and safety and reliability. The R&D work concentrates on these issues. The development of martensitic steels, including MANET, is part of a separate programme. Breeder ceramics and beryllium irradiations have been performed so far under conditions which do not cover the peak values projected for the DEMO blankets. Further irradiations in thermal reactors and in fast reactors, especially for beryllium, are required. Effective tritium control requires the development of permeation barriers and/or of methods of oxidation of the tritium in the main helium cooling system. First promising results have been obtained also in the field of mechanical effects from plasma disruptions and safety and reliability. However, further work is required in the reliability field and to validate the codes for the calculations of the plasma disruption effects.

Extraction and Powder Metallurgy of Beryllium

The paper presents beryllium extraction process being operated at the beryllium pilot plant of Bhabha Atomic Research Centre. The metal is produced by magnesium reduction of anhydrous beryllium fluoride. Reduction process yields beryllium metal in the form of pebbles. Metal pebbles are refined by vacuum induction melting. Vacuum cast ingot is further processed through powder metallurgy to produce vacuum hot pressed beryllium blocks. These blocks are precision machined, in the Beryllium Machining Facility, to various components for space applications.

Assessment of water-cooled beryllium components for plasma-facing applications

The successful operation of JET in the presence of beryllium target plates has been limited by overheating of the plates. Active cooling of these components is investigated as a way of solving this problem. Tests carried out at JET suggest that water-cooled components comprising beryllium clad to heat sinks have a safe margin of steady state operation up to 12 MW/m 2. Results from the irradiation of inertially cooled beryllium blocks suggest a limit of ∼ 12 MW/m 2 to be sustainable by S65B beryllium without low-cycle fatigue cracking obtained for initial temperatures > 200°C.

Analysis of the Thermomechanical Contact between Beryllium Blocks and Stainless Steel Plates in ITER Design

The thermomechanical contact between beryllium cubes and Type 316 stainless steel was analyzed for various values of applied pressure normal to the interface. If the authors neglect the influence of gap on the interface resistance, finite element analyses show that a simple one-dimensional analysis can lead to serious underestimation of the maximum temperature of the beryllium. A two-dimensional analysis underpredicts the maximum gap created at the interface, compared with a full three-dimensional analysis. Thus, it also significantly underpredicts the maximum temperature of the beryllium. 8 refs., 15 figs.