Search

A concept of a sodium cooled fast spectrum materials testing reactor is presented as one of the potential design configurations to achieve the performance versatility with respect to supporting a range of testing environments from current thermal-spectrum LWRs to Generation IV advanced fast-spectrum reactors. The objective is accomplished by designing a core with region-wise varying neutron spectra. This paper presents the design details and explores spectral variability examining feasibility to maintain thermal and fast energy spectra in the system. The core assemblies are designed using the EBR-II assembly as a prototype while implementing certain design modifications to meet performance needs. The fuel is 19.9%-enriched U-10Zr. Compare to EBR-II, the proposed core has larger pins, a taller active core, and thick reflector instead of depleted uranium breeding blankets. The MgO pins are analyzed as the reflector material to meet the needed versatility requirements. The design effort successfully concluded yielding a feasible 600 MW th -configuration with sufficient neutron fluxes in material testing locations. A fast flux greater than 5.6E15 n/cm 2 -s is maintained over the core lifetime in the central irradiation position. The volume of 63,000 cm 3 with a fast flux of ∼ 4.2E15 n/cm 2 -s or greater is provided for materials irradiation options. The core has regions of high magnitude thermal flux in a large graphite region at the core periphery. Spectral variability, design choices and reactor physics characteristics are discussed to demonstrate the concept viability.

A dedicated neutronics computational scheme is a crucial tool in the preparation and the post-analysis of an irradiation experiment, as it defines the key parameters of the experiment. The complex and very dynamic environment of a Materials Testing Reactor (MTR) core, on the other hand, represents a challenging object for neutronics calculations. Within the framework of the HELIOS experiment conducted at the High Flux Reactor (HFR) in Petten (The Netherlands) in 2009 under the European Project EUROTRANS, a new neutronics scheme has been developed with the objective of finding the optimum balance between accuracy and calculation performance. This scheme is based on the Monte-Carlo technique and employs a two-level approach by separating the problem into overlapping spatial and temporal domains. The underlying physical phenomena are presented, together with elements of validation which gives a further insight into the effects governing the neutron fluxes in the core.

Reactor pressure vessel (RPV) steels for pressurized water reactors (PWRs) with bulk P contents ranging from 0.007 to 0.012wt.% were subjected to neutron irradiation at fluences ranging from 0.3 to 1.2 × 1020 n/cm2 (E > 1 MeV) in PWRs or a materials testing reactor (MTR). Grain-boundary P segregation, which was analyzed using Auger electron spectroscopy (AES) on intergranular facets, increased with increasing neutron fluence. A rate theory model based on four diffusion-reaction equations for substitutional P atoms, octahedral interstitial P atoms, vacancies, and self-interstitial atoms was also used to simulate the increase in grain-boundary P segregation for RPV steels with a bulk P content up to 0.020wt.%, using parameters optimized by the present AES data. The increase in grain-boundary P segregation in RPV steel with a bulk P content of 0.015wt.%, which is the maximum P concentration in RPV steels used in Japanese nuclear power plants intended for restart, was estimated to be less than 0.1 in monolayer coverage at 1 × 1020 n/cm2 (E > 1 MeV). A comparison of the PWR data with the MTR data, including that from the literature, showed that neutron flux had no effect upon grain-boundary P segregation for A533B steels. The relationships of the ductile-brittle transition temperature (DBTT) shifts to grain-boundary P segregation and to yield strength were also discussed. A linear relationship between the yield strength and the DBTT shift with a slope of 0.63 was obtained for RPV steels with a bulk P content up to 0.026wt.%, which is higher than that of most U.S. A533B steels. It is concluded that the intergranular embrittlement is unlikely to occur for RPV steels irradiated in PWRs.

A prototype 3 He-based Passive Neutron Albedo Reactivity (PNAR) counter was developed and tested at Los Alamos National Laboratory (LANL) in collaboration with the Korea Atomic Energy Research Institute (KAERI) to measure the fissile content in electrochemical recycling (ER) product materials. The counter consists of 16 3 He cylindrical gas-filled proportional counters at 4 atm of pressure embedded in high-density polyethylene. In this work, experimental measurements were performed at LANL to characterize the performance of the PNAR counter using surrogate materials for the uranium metal ingot. The purpose of these experiments was to: 1) measure the operating and calibration parameters of the PNAR counter (e.g. efficiency profiles, coincidence gate fractions, die-away time) and 2) evaluate the accuracy and sensitivity of the PNAR method and the time correlated induced fission (TCIF) method for quantifying the 235 U mass in PWR fresh LEU fuel rods and Materials Testing Reactor (MTR) HEU fuel plates. A small 244 Cm reference source (13,373 n/s) was placed in the center of the fuel rods and fuel plates to simulate spontaneous fission from sub-ppm (parts per million) levels of Cm contamination in the U ingot. In order to compare the relative accuracy of the PNAR and TCIF methods for quantifying 235 U mass, calibration curves were generated for the net doubles rate and the doubles Cd ratio using the Deming software. The results from this experiment will be used to obtain a better understanding of the sensitivity of the PNAR and TCIF methods for samples with low neutron multiplication. Furthermore, this experimental measurement data will also help inform safeguards research and development (R&D) efforts on the viability of nondestructive assay (NDA) techniques and detector designs for quantifying fissile content in ER product materials. Future work will include performing measurements with the PNAR counter on small samples of U/TRU materials.

The defect structures in Ni, Fe–15Cr–16Ni and Ti-added modified SUS316SS (modified SUS316) were examined after neutron irradiation below 0.3 dpa by the Japan Materials Testing Reactor and Belgian Reactor 2 to compare their defect structural evolution. The growth behaviour of interstitial-type dislocation loops (I-loops), stacking fault tetrahedra (SFTs) and voids was found to be quite different among these specimens. I-loops developed at lower temperatures in Ni than in Fe–15Cr–16Ni and modified SUS316, and more swelling occurred in Ni than in Fe–15Cr–16Ni. Finally, there were no voids in modified SUS316. These results were analysed in terms of the I-loop energy. A large discrepancy was found between the analytical results and experimental observations for Ni and modified SUS316, which suggests the formation of unfaulted I-loops directly from collision cascades. The growth of SFTs was detected in Fe–15Cr–16Ni and modified SUS316, and can be explained by a change in the dislocation bias of SFTs resulting f...

A small-sized motor with a resistance to neutron and gamma-ray radiation was developed. This motor was rotating normally at a gamma-ray dose of 7 × 10 8 Gy, a fast neutron ( E > 1 MeV) fluence of 2 × 10 21 m −2 and a thermal neutron ( E 22 m −2 . The gamma-ray dose for this motor was about 700 times as high as the specification of a commercial motor. Much progress in radiation resistance was attained by using radiation-proof materials without organic lubricant. It became clear that this small-sized motor is able to be used for in-pile tests of a fusion blanket using the Japan Materials Testing Reactor (JMTR).

Radiation induced changes in electrical properties of calcium-zirconium-indium oxide ceramics (CaZr0.9In0.1O3-d) were investigated using a fission reactor of Japan Materials Testing Reactor (JMTR). Electrical conductivity of CaZr0.9In0.1O3-d under the reactor irradiation increased by the Radiation Induced Conductivity (RIC), Radiation Induced Electromotive Force (RIEMIF)and Radiation Induced Electrical Degradation (RIED), caused by the electronic excitation. The RIC and the RIEMF decreased gradually for the constant value of the reactor power with the increase of irradiation time. Moreover, it was confirmed by Post-irradiation Examination (PIE) experiments that the initial electrical and thermal properties were completely degraded by the irradiation. These radiation effects are attributed to point defects or their clusters produced by atomic displacements caused by neutron collisions and also by the radiolysis.

A binary pebble bed of lithium titanate (Li 2 TiO 3 ) was irradiated in the Japan Materials Testing Reactor (JMTR), and its tritium recovery characteristics bed was studied under pulsed neutron operations. The temperature at the outside edge of the pebble bed increased from 300 to 350 °C immediately after the window of hafnium (Hf) neutron absorber was turned toward the reactor core, while the tritium recovery rate increased gradually. The ratio of tritium recovery rate to generation rate at the high-power, ( R / G ) high , approached the saturated value of unity at about 20 h of operation. Overall tritium recovery behavior under the pulsed operation was similar to that under the steady state power operation. An estimated time constant of about 3 h for the tritium recovery was much longer than the thermal time constant of about 100 s.

Radiation effects in components and materials will be one of the most serious technological issues in nuclear fusion systems realizing burning-plasmas. Especially, diagnostic components, which should play a crucial role in controlling plasmas and understanding the physics of burning-plasmas, will be exposed to high-flux neutrons and gamma rays. Dynamic radiation effects will affect the performance of components substantially from the beginning of exposure to radiation environments, and accumulated radiation effects will gradually degrade their functioning abilities in the course of their service. High-power-density fission reactors will be the only realistic tools to simulate the radiation environments expected to occur in burning-plasma fusion machines such as the International Thermonuclear Experimental Reactor (ITER), at present. Some key diagnostic components, namely magnetic coils, bolometers, and optical fibres, were irradiation-tested in a fission reactor, to evaluate their performances in heavy radiation environments. Results indicate that ITER-relevant radiation-resistant diagnostic components could be developed in time, although there are still some technological problems to be overcome.

In order to investigate effects of neutron irradiation on fatigue properties of nuclear materials, a remote-controlled high temperature fatigue test machine was developed at the hot laboratory of the Japan Materials Testing Reactor (JMTR) in the Japan Atomic Energy Research Institute (JAERI). A small-sized fatigue specimen having double blades to measure strain with a laser extensometer was designed for this machine. A strain amplitude in fatigue tests of a completely reversed push-pull type using a triangular wave was controlled with an accuracy of ±3% of the total strain range during test. Low cycle fatigue tests of type 304 stainless steel irradiated in JMTR at 823 K up to a fast neutron fluence of 1×1025 n/m2 (E>1 MeV) were performed in total strain ranges of 0.7–1.4% at 823 K using the designed small-sized specimens.

To investigate the microstructure evolution during temperature change irradiation, Ni and its binary alloys were irradiated in the Japan Materials Testing Reactor to doses up to 0.46 dpa under two cycles of temperature change of 533/693 K, i.e., 533 K, then 693 K, then 533 K, then 693 K. The irradiation dose at each temperature stage was about the same. The specimen sets were pulled out five times during irradiation. In the beginning of the first irradiation at the high temperature of 693 K after irradiation at 533 K, the densities of interstitial type dislocation loops and voids decreased in pure Ni, Ni–2at.%Cu and Ni–2at.%Ge alloys. In the second irradiation at 533 K, however, the formation of dislocation loops and voids was promoted. The void density continued to increase even after the second irradiation at 693 K. On the other hand, in Ni–2at.%Si and Ni–2at.%Sn alloys, where the formation of voids was suppressed by addition of the minor elements Si or Sn, the microstructure evolution was almost independent of irradiation temperature.

The study of neutron irradiation damage is inevitable for the development of fusion reactors such as the International Thermonuclear Experimental Reactor (ITER). In general, it is difficult to simulate the characteristics of irradiation damage, such as the relationship between helium production and displacement damage, expected in fusion reactors by using the existing test reactors. This is due to the neutron spectrum of test reactors, which are more thermalized than expected in fusion reactors. Despite such limitation, irradiation damage expected to occur in fusion reactors can be simulated in test reactors by spectrum tailoring techniques and/or a combination of irradiation at different conditions, so called irradiation–coupling. In Japan Atomic Energy Research Institute (JAERI), the technique of re-irradiation for irradiated specimens was developed to realize irradiation–coupling as stated above. The key technology in establishing such a technique is the development of a specially designed coupling capsule which can be assembled and loaded with irradiated specimens by remote handling devices in the hot cell. JAERI has recently developed the coupling capsule and successfully completed its irradiation in the Japan Materials Testing Reactor (JMTR).

The effect of alloying elements, Si (-5.8%: the volume size factor in Ni), Ge (+14.76%) and Sn (+74.08%), on void swelling in neutron irradiated Ni at 573 K was studied by transmission electron microscope (TEM) observation and positron annihilation lifetime measurement. Neutron irradiation dose was changed widely from 0.001 to 0.4 dpa using two reactors, the Kyoto University reactor (KUR) and the Japan materials testing reactor (JMTR). Voids were observed in pure Ni by TEM even after very small irradiation dose of 0.001 dpa. With increasing dose, the density of voids did not change much while their size increased. The same tendency was observed in Ni-2at.%Ge. In Ni-2at.%Sn and Ni-2at.%Si, however, no voids were observed by TEM at a damage dose of 0.4 dpa. But positron lifetime measurement revealed the existence of microvoids at a medium dose of irradiation. When irradiation dose increased to 0.4 dpa in Ni-2at.%Si and 0.13 dpa in Ni-2at.%Sn, their existence was not detected. Suppression of microvoids in these alloys is discussed from the standpoint of solute point defect interactions.

This paper consists of two parts. In part 1, the experimental results of damage evolution of neutron-irradiated Cu and Ni are described. In part 2, results of computer simulations are described with linkage of experimental data to explore the atomistic process of damage evolution. To study experimentally the atomistic processes of damage evolution in neutron-irradiated Cu and Ni in part 1, we prepare two types of specimens for both metals. One is as-received specimen from manufacturer. Another is a residual-gas-free specimen which is prepared by melting as-received metals in highly evacuated vacuum at 10−5 Pa. Specimens are irradiated with fission neutrons in the temperature-controlled-irradiation capsule at JMTR (Japan Materials Testing Reactor). TEM (Transmission Electron Microscope) observation shows that the dislocation structure is developed by the aggregation of interstitial clusters in irradiated metals. It is found that the number density of void which are observed in specimens, both as-r...

Mg-Al spinel ceramics with different stoichiometry (Al 2 O 3 /MgO=n; n = 1.00, 1.01, 1.10 and 1.48) were irradiated with fast neutrons up to 3.8 x 10 23 n/m 2 (E > 0.1 MeV) at 170°C, and 5.7 x 10 23 n/m 2 at 190°C in the Japan Materials Testing Reactor. There was no obvious difference in property change between the two irradiation conditions. The amount of macroscopic length change due to these irradiations increased almost linearly with increasing Al 2 O 3 content. These specimens were isochronally annealed up to 1000°C, During the annealing, macroscopic length of the specimens started to decrease at around 200°C and reduced monotonically up to 550°C, After that, the length increased slightly up to 650°C and then decreased again up to 800°C, showing a small peak. These final values corresponded to the length of the specimens before the irradiation. The shrinkage in length between 200-550°C and 650-800°C may be attributed to the recombination of Frenkel pairs of anions, and the expansion in length between 550-650°C to the order-disorder transition of cation distribution.

Neutron irradiation effects on a Ti–Ni shape memory alloy (SMA) are not understood regarding the engineering behavior of tightening and connecting core parts. Irradiation response on shape memory capabilities for Ti–Ni SMA couplers was investigated by measuring the inner diameter of a SMA coupler. Ti–Ni SMA couplers were irradiated to a fast neutron fluence ( E > 1 MeV) of 1.4 × 10 25 m −2 at 323 and 503 K in the Japan Materials Testing Reactor of JAERI. The inner diameter versus temperature curves showed a large hysteresis for the SMA coupler irradiated at 323 K. At a neutron fluence above 10 25 m −2 , the hysteresis after 323 K irradiation increased by 50% more than that after 503 K irradiation. The maximum amount of reversible stroke of the irradiated two-way Ti–Ni SMA coupler during heating and cooling was reduced to 70% that of an unirradiated one and the two-way Ti–Ni SMA coupler is more resistant to neutron irradiation than the one-way Ti–Ni SMA coupler.

A beryllium irradiation study with the Japan Materials Testing Reactor of JAERI is reviewed from the point of view of tritium release, thermal properties, mechanical properties and reprocessing technology by using the neutron irradiated sample with a total fast neutron fluence up to 10 21 (n cm −2 ). It becomes clear that problems arise for each of the properties, for example, effects of the surface oxide layer, impurities, grain size, and/or swelling. Moreover, future plans of beryllium irradiation studies are discussed in this paper.

TiNi and TiPd shape memory alloys (SMA) reveals good workability and shape memory properties and these SMAs seem to be one of hopeful functional materials in a severe irradiation field. The effects of irradiation induced residual strain on the deformation processes of TiNi and TiPd SMAs after neutron irradiation with fluences ( E > 1 MeV) up to 3.9 × 10 24 m −2 at Japan Materials Testing Reactor (JMTR) were investigated by the remote controlled X-ray diffraction measurement. Residual strains of TiNi SMAs took place over damage of 0.1 dpa and the strains were not completely removed by post-irradiation annealing at 473, 523 and 573 K. On the other hand, residual strains of TiPd SMAs scarcely occurred after irradiation, and furthermore, the TiPd SMA seems to be an irradiation-resistant material. This may be explained by a difference between the irradiation response of a parent phase and that of a martensitic phase to neutron irradiation. Deformation processes of SMAs are associated with stress fields generated by irradiation in a parent phase with B2 type ordered structure or in a martensitic phase with 2H and 9R type close-packed structures.

High thermal diffusivity and insulating aluminum nitride (AIN) ceramic was neutron-irradiated in the Japan Materials Testing Reactor. Fast neutron fluence and irradiation temperature of the specimens were 1.8 × 10 23 n/m 2 at 100°C and 5.3 × 10 24 n/m 2 at 470°C ( E > 0.1 MeV), respectively. Change in thermal diffusivity due to isochronal thermal annealing up to 1400°C was measured by the laser-flash method. As compared with macroscopic length change, it was concluded that thermal diffusivity of irradiated AIN was controlled by concentration of irradiation-induced vacancies.

Tritium release apparatus with the function of pulse mode heating was developed by using the infrared ray furnace to demonstrate the pulse mode heating of tritium breeder blanket for the fusion reactor. This apparatus was installed in the glove box of the beryllium PIE facility that constructed the hot-laboratory of Japan Materials Testing Reactor. The performance of this apparatus is that the minimum time of rapid heating up to 1015 {degree}C is about 119 s and maximum heating rate reached at 1000 {degree}C/min. The maximum temperature depends on the crucible materials because of the differences for infrared ray absorption. The conversion efficiency of the gaseous water by ceramic electrolysis cell is above 99.99%. The pulse mode heating of the tritium breeder and neutron multiplier materials of the blanket could be demonstrate by using this apparatus. 7 refs., 7 figs., 1 tab.

MgO·3Al2O3 single crystals were irradiated with neutron fluences of 8.3 × 1022 n/m2 at 100°C and 2.4 × 1024 n/m2 at 470°C (E > 1.0 MeV) in the Japan Materials Testing Reactor. The Knoop microhardness of several orientations on the (100) plane of both the irradiated and unirradiated crystals were measured with different indentation loads. The change in hardness profile of the crystals was almost the same after the two irradiation conditions. The hardness increased by 4–15% because of the irradiations depending on the crystallographic orientation, the larger change being observed at orientations between the (001) and (011) directions. While both the {111} and {110} slip systems are simultaneously active in the unirradiated MgO·3Al2O3, the {111} system may be the dominant slip system in the neutron-irradiated crystals. It is concluded that the restriction of the {110} slip system is caused by irradiation-induced interstitial ions.

Thermal diffusivity of two kinds of two-directional carbon-carbon composites were studied after neutron irradiation at temperatures 640–1070°C to a maximum neutron fluence 1.98 × 10 25 n/m 2 ( E > 29 fJ , 0.18 MeV) in the JMTR (Japan Materials Testing Reactor). The decrease of thermal diffusivity was observed after irradiation, and its significant decrease occurred at low neutron fluence, followed by gradual decrease with neutron fluence. The changes in thermal diffusivity, which were normalized to unirradiated values, were large at low measurement temperatures and decreased with temperature and also increased with neutron fluence. The changes in thermal conductivity related to thermal diffusivity are also discussed.

Magnetic probes are installed on the first wall of a fusion reactor to measure the magnetic field. These magnetic probes are exposed to the neutrons generated by the D-T reaction and γ-rays. However, electrical properties and durability of the magnetic probe under irradiation have not yet been satisfactorily investigated. Therefore, irradiation tests have been performed for two types of magnetic probes at the JMTR (Japan Materials Testing Reactor). It has resulted from these irradiation tests that the inductance of the magnetic probe with ceramic coated platinum wire is stable below about 550°C, and the inductance of the magnetic probe with MI-cable of 0.5 mm diameter has a temperature dependency and maintains reproducibility even above 550°C.

The tensile properties of reduced-activation martensitic 9Cr-IWVTa and 9Cr-3WVTa steels for fusion reactors were investigated over the temperature range from room temperature to 873 K after neutron irradiation in the Japan Materials Testing Reactor at 538 K to fast neutron fluences of 2×10 23 and 3×10 24 n/m 2 (E>1 MeV). A conventional 9Cr-1MoVNb steel was also examined for comparison. The irradiation caused an increase in yield stress and a decrease in total elongation. Irradiation-produced defect clusters could not be seen directly by transmission electron microscopy. The increase in yield stress caused by irradiation Δσ y was proportional to the 1/4 power of the neutron fluence

Aluminum nitride was neutron-irradiated in the Japan Materials Testing Reactor up to 5.2×10 24 n/m 2 ( E > 1.0 MeV). The specimens were isochronally annealed under a vacuum up to 1400°C. The macroscopic length and lattice parameters of both a - and c -axis of the specimen irradiated to 8.3 × 10 22 n/m 2 at about 100°C decreased concurrently with an increase in annealing temperature from 100°C, and returned to the pre-irradiated values after annealing at 1100°C. The macroscopic lengths of the specimens irradiated to 2.4×10 24 n/m 2 at 470°C and 5.2×10 24 n/m 2 at 785°C, both containing dense interstitial dislocation loops, were not significantly influenced by annealing up to 800–900°C. Then they decreased rapidly with increasing annealing temperature up to 1400°C. The change in c -axis lattice parameter by annealing was larger than that of a -axis above 900–1000°C, and finally reduced to the pre-irradiated values after annealing at 1400°C.

Tensile properties of the reduced activation martensitic steels 9Cr-1WVTa and 9Cr-3WVTa were investigated over the range from room temperature to 873 K after irradiation in the Japan Materials Testing Reactor at about 538 K. Modified 9Cr-1MoVNb steel was also examined for comparison. The increase in yield stress due to irradiation measured at room temperature was smaller in the 9Cr-1WVTa and 9Cr-3WVTa steels than in conventional 9Cr-1MoVNb steel and was smaller in the low-W steel than in the high-W steel. With increasing test temperature, the irradiation hardening exhibited a further increase at around 573 K and then decreased to zero over the range 673 to 873 K. The increase in irradiatio hardening at around 573 K was believed to be due to vacancy-carbon complexes. The present results were also compared with earlier irradiation hardening experiments.

Magnetic flux densities for neutron irradiated specimens of austenitic stainless steels, SUS304 and Fast Breeder Reactor grade type 316 (316FR), were measured by using a flux gate (FG) sensor to investigate the nondestructive evaluation method of irradiation damage parameters, dose and He content. Specimens were irradiated in each one of the experimental fast reactor JOYO, the Japan Materials Testing Reactor, and the Japan Research Reacter-3M (JRR-3M), or in both of JRR-3M and JOYO (coupling irradiation). Irradiation in various reactors and the coupling irradiation provided irradiation conditions which could be hardly obtained by irradiation in a single reactor. The range of dose, He content and irradiation temperature of the neutron irradiated samples studied in this paper were 0.01–30 displacement per atom (dpa), 1.0–17 appm and 470–560 °C, respectively. Magnetic flux density increased with dose although there may be a threshold dose for magnetic property to change between 2 and 5 dpa for 316FR. This result shows the possibility of nondestructive evaluation of dose by measuring magnetic flux density by an FG sensor. On the other hand, magnetic flux density did not depend on He content.Copyright © 2009 by ASME

Emission monitoring of radionuclides has been performed since 1959 at the Studsvik site at the Baltic coast in the county of Södermanland, 80 km south of Stockholm. Several small nuclear reactors have been used at the site. Today only one 50 MW high-flux pool-type materials testing reactor (MTR) and one 1 MW low-power pool reactor are used. In the R2 reactor, testing of nuclear reactor fuel, production of radioisotopes for medical purposes and neutron transmutation doping of silicon are performed. A facility for boron neutron capture therapy (BNCT) for medical treatments at the R2-0 reactor will be in operation by the year 2000. The University of Uppsala has a department of neutron research at the reactor. Other types of nuclear work done at Studsvik are decontamination, incineration, melting and recycling of low-level waste. Yearly emissions are much smaller than the limit, 0.1 mSv per year to the critical group, set by the Swedish authority.

A very detailed 3-D MCNP-4B model of the BR2 reactor was developed to perform all neutron and gamma calculations needed for the design of new experimental irradiation rigs. The Monte Carlo model of BR2 includes the nearly exact geometrical representation of fuel elements (now with their axially varying burn-up), of partially inserted control and regulating rods, of experimental devices and of radioisotope production rigs. The multiple level-geometry possibilities of MCNP-4B are fully exploited to obtain sufficiently flexible tools to cope with the very changing core loading.

Beryllium samples have been irradiated in BR2, the materials testing reactor of the nuclear research centre in Mol, Belgium, up to fluence values of 4.1022 n/cm2 (E>lMeV). The helium production and release characteristics, as well as the tritium content have been determined as a function of the neutron fluence. The induced swelling of the beryllium and the enhancement of this swelling due to annealing have also been evaluated. The paper describes the experimental methods applied and the results obtained. In parallel, the helium and tritium productions are being calculated and correlated with the neutron fluences, and compared with the measured values. The work is performed in the framework of the Fusion Blanket Programme of the European Commission.

Thermal shock tests of neutron irradiated carbon fiber reinforced carbon composites (CFCs) were started with OHBIS to study the neutron irradiation effects on the erosion behavior. Specimens were one directional, two directional and three directional CFCs. These specimens were irradiated with a total neutron fluence of 0.3–0.4 dpa at 556-594 K using Japan Materials Testing Reactor. Heat load conditions were 20 MJ m-2 (800 MW m-2 × 25 ms and 500 MW m-2 × 40 ms). It was clear that the weight loss of the neutron irradiated specimens increased with neutron fluence, and that the erosion area of the neutron irradiated specimens were broader than that of the un-irradiated specimens.

The branching ratio, β − / EC = 5·1 ± 0·1, for 242 m Am decay has been measured by determining the relative amounts of the decay products, 242 Cm and 242 Pu formed. The 16 hour 242 m Am was produced by neutron irradiation of 241 Am in the Materials Testing Reactor. A pile neutron cross-section of 620 ± 65 barns for the reaction 241 Am(n, γ) 242m Am was measured.

The gamma-ray dose rate and the fast neutron flux have been measured in air from a source of water shortly after irradiation in the Materials Testing Reactor. Gamma rays from N16 and neutrons from ...

Measurements of the partial differential neutron scattering cross sections for room-temperature propane gas are reported. These measurements were made at incident energies of 0.0101, 0.0254, 0.0736, and 0.102 ev at seven scattering angles between 16.3\ifmmode^\circ\else\textdegree\fi{} and 84.7\ifmmode^\circ\else\textdegree\fi{} using the Materials Testing Reactor phased chopper velocity selector. The data are converted to the scattering-law presentation and compared with three theoretical calculations: (a) The ideal gas, using an effective mass obtained from an average of the mass tensors for the three types of H atoms in propane, gives poor agreement. (b) The Krieger-Nelkin approximation, which includes the effect of zero-point vibrations, gives limited agreement for energy transfer less than $0.5{k}_{B}T$ at intermediate momentum transfers. At large momentum transfers where vibrational effects become important it underestimates the cross section. (c) A modification of the Krieger-Nelkin theory that includes the effects of single-quantum transitions from the three lowest vibrational states gives better agreement. The discrepancies still present at large momentum and energy transfers are attributed to an uncertainty in the methyl-group barrier height for the three lowest energy modes, to the harmonic oscillator approximation for these modes, and to the approximate molecular orientation averaging used in the calculation.

Small pieces of a beryllium shim rod from the Materials Testing Reactor have been heated and changes in microstructure, volume and mechanical properties observed. The beryllium which was estimated to have received a dose of about 7.6 × 1021 fast neutrons/cm2 contained about 10 cm3 at N.T.P. of helium per cm3 and this precipitated as fine gas bubbles during heating for an hour at temperatures above 600° C. These bubbles grew, particularly at the grain boundaries, during further heating, until at 1100° C they were about 10−3 cm diameter. The precipitation and growth of the bubbles produced an increase in the volume of the samples, commencing at about 600° C and eventually reaching 30 % at 1000° C. The irradiated beryllium is very brittle, even above room temperature, and this may be aggravated by the large gas bubbles lying on the grain boundaries. The observations are discussed in relation to those already reported for beryllium containing large local concentrations of helium injected with a cyclotron.

The burst shapes and transmission probabilities for neutrons passing through a phased chopper neutron velocity selector of the type now being used at the Materials Testing Reactor have been calculated using an IBM-650 computer. Choppers with three types of slits (straight, parabolic, and circular) are considered and burst shapes are presented that show the operating characteristics of such velocity selectors. A function that gives a good fit to many of the burst shapes is presented. The results verify that a circular slit chopper can be used in place of the more exact parabolic slit chopper with only minor distortions in the operating ranges of interest.

The Materials Testing Reactor (MTR) velocity selector consists of a pair of Fermi-type phased choppers and a pair of rotating collimators. The four rotors are driven at identical speeds up to 5225 rpm by four hysteresis synchronous motors. With the appropriate angular phase difference established between the choppers and the collimators, bursts of monoenergetic neutrons are selected from the MTR's Maxwellian flux. Phase instability broadens the initial neutron burst by only 12%, as demonstrated by actual measurements of time distributions. The monochromatic neutrons produced by the selector are used in measuring inelastic scattering cross sections by means of time-of-flight equipment. Intensities per time channel are sufficient to give reasonable statistics in a 24 hour run. Wavelength resolutions between 2% and 8% are realized. The paper describes the layout and characteristics of the velocity selector in detail, and discusses problems associated with backgrounds and time and intensity calibrations. Inelastic scattering data taken on samples of methane gas and on a beryllium single crystal are displayed and analyzed to show the counting rates and background levels achieved with the selector, and to demonstrate the versatility possessed by this type of monochromator.

With the availability of a high-intensity neutron beam from an underground nuclear explosion, it is possible to measure short half-life isotopes a few hours after final separation and before the build-up of decay products. The decay-product contaminants may be considered constant during the measurement as the measuring time is only a few msec. A hot-sample changer was built in conjunction with Los Alamos Scientific Laboratory to measure neutron cross sections using time-of-flight methods on samples of high activity and short half-lives (a few days). Samples are produced by neutron irradiation in the Materials Testing Reactor (MTR) and prepared in hot cells. They are flown to the Nevada Test Site (NTS) for insertion into the sample-changer cask a few hours before the experiment. The sample is moved into beam position 15 sec before the device is detonated and is returned to the charger cask immediately thereafter. The capture cross section of the nuclide 147 Pm was measured in the spring of 1967 and served as a “dry run” for the sample 233 Pa, which is being measured in a current experiment.

The thermal neutron flux in a water medium from a source disk containing water irradiated in the Materials Testing Reactor has been measured using a BF3 detector. Neutrons from N17 and photoneutron...

A high-intensity beam of 2 keV neutrons having a full width at half maximum of 800 eV has been produced by placing a scandium filter in the Materials Testing Reactor. Preliminary studies have been carried out on several samples showing how the total neutron cross section at 2 keV changes as a function of sample temperature. These studies show that certain materials, particularly those with sharp resonance structure at 2 keV, exhibit marked changes in resonance self-shielding with temperature, as a result of the Doppler broadening of these resonances.

The low-energy total neutron cross section of Pu/sup 241/ has been measured from 0.02 ev to 2 kev using the Materials Testing Reactor fast chopper. Curves of the total cross section are given and pronounced asymmetries are observed in several of the resonances below l2 ev. An estimate of the neutron strength function obtained from the average cross section in the kev energy region is (0.85 plus or minus 0.15) x 10-/sup -4/. (auth)

The partial differential cross sections for inelastic scattering of neutrons from samples of cold gaseous methane (T=125°K), liquid methane (T=99°K), and solid methane (T=83°K) were measured with the materials testing reactor (MTR) phased chopper velocity selector. These results are compared to the theory reported by Griffing for methane gas, which fits the cold‐gas data very well over the entire range of the experiment and also fits the liquid and solid data at the large momentum transfers. The results of these comparisons indicate that the rotations are free in all three phases and that, in the liquid and solid phases, there are oscillations of the center of mass of the molecule with dominant transitions having an energy change near 0.017 eV. The range of agreement of the gas theory with the liquid and solid data indicates that the liquid and solid act according to the gas model out to a time of roughly 5×10−13 sec. The similarity of the liquid and solid data over the range of the experiment suggest fur...

A new spontaneous-fission activity has been observed in a fermium fraction separated from 120 mg of mixed curium isotopes which had been subjected to an integrated flux of 2·4 × 10 22 neutrons per cm 2 at the Materials Testing Reactor, Idaho Falls Idaho (MTR). A least-squares analysis of the spontaneous-fission decay data yielded a half-life of 10·8 days with a standard deviation of about ( +2.2 −1.8 days. An estimated half-life of (11 +10 −6 days is reported here. The mass assignment is uncertain, but limited to A = 258 or 257.

Engineering materials irradiations form a substantial program at Oak Ridge National Laboratory (ORNL), relying mainly on the High-Flux Isotope Reactor (HFIR) and the Oak Ridge Research Reactor (ORR). The HFIR neutron flux is high, but the reactor was not designed for materials irradiations; the lack of instrumented positions in the target region and the small size and limited number of positions in the high-flux region immediately outside the control plates pose particular disadvantages for this kind of research. Relatively minor modifications to the HFIR, at a cost of less than $2 million, would give it a first-class materials irradiation capability, capable of undertaking most of the experiments now performed at the ORR in one-half to one-third the time. The lack of comprehensive neutronics calculational capability at ORNL is a major drawback for experimenters and for the reactor operators, making it difficult to assess the feasibility of proposed new experiments or reactor improvements. The Operations Division should acquire the necessary computational capability, making use of new computational techniques and the newly available, more powerful computers as necessary. As a long-term prospect, when the proposed Center for Neutron Research (CNR) is build and the HFIR is no longer needed for neutron-scattering workmore » and isotope production, the HFIR core could be replaced by a general purpose materials testing reactor core; the result would be the world's finest materials testing facility for less than one-tenth the cost of a new reactor. The value of such a scheme will depend on the national need for materials irradiation testing and isotope production at the time that the CNR is commissioned.« less

A program was directed toward development of a thermopile-type neutron- flux-sensitive meter of small physical size and having a fast response time (25 msec) and a high signal output. The meter configuration selected consists essentially of a metallic heat sink, a thermopile, and a thin layer of neutron- sensitive material in series. When the meter is exposed to a neutron environment, heat generated in the neutron-sensitive material flows across the thermopile unit to the heat sink. The signal output generated by the thermopile is proportional to the heat flow, which is in turn proportional to the neutron flux. Signal-output characteristics of meters of this configuration have been verified by experimental testing in the Battelle Research Reactor, the General Electric NTR Reactor, and the Materials Testing Reactor (Idaho Falls). A practical neutron meter, in which the sensing element measures approximately 1/8 in. x 1/2 in. x 1 in., with a time constant of less than 25 msec and a millivolt- level signal output at a thermal-neutron flux of 10i/sup 12/n/cm/sup 2/-sec was achieved, thus substantially fulfilling the program objectives. (J.R.D.)

Results are given of some experiments to measure the light output from NE 102 and NE 421 scintillators under neutron radiographic conditions. These are then used to interpret an experiment in which an image intensifier was used to provide immediate radiography by direct viewing of a scintillating plate placed in the Materials Testing Reactor neutron radiography unit at Dounreay.

The long‐wave spectrum of a neutron beam from the Be reflector of the Materials Testing Reactor (MTR) has been measured from 3 A to 10 A by a total reflection technique. The spectrum was found to vary as λ−3.9.