14 MeV中子辐照钒引起的长寿命放射性

用HPGe探测器测量了钒以及杂质由14MeV中子活化引起的放射性,得到了50V(n,α)和51V(n,α)反应截面比值在14MeV处为0 584,并由此得到了50V(n,α)反应截面为(8 7±0 7)×10-27cm2。测量了钒中杂质成分被活化后其长寿命放射性核素的半衰期。     

Neutronics Experiments of D-T Fast Neutrons by Application of Associated Particle Method

A probabilistic method is proposed that will permit in the design stage to estimate quantitatively the likelihood with which any or all design criteria applicable to a nuclear reactor are actually satisfied after its construction. The method is trially applied to the core reactivity balance problem of the experimental Very High Temperature Reactor, and calculations are performed on the probability with which a design study core will, upon construction, satisfy design criteria concerning (a) one rod stuck and (b) startup margin. The method should prove useful in making engineering judgments before approving reactor core design.     

The Effect of Cylindrical Holes in a Moderator on the Passage of Virgin Neutrons from a Point Source

The mean free path and the mean square free path of 14 MeV neutrons, generated isotropically from a d-t point source located at the center of the moderator region of a square lattice composed of parallel circular cylindrical holes in water, are calculated for various void ratios. Very interesting inequalities between the axial and transversal components of the mean free path and the mean square free path are found, which contradict with the results of Behrens for homogeneous neutron sources. We may conclude that the effects on the free path lengths of the position of the point source located in an anisotropic and heterogeneous medium cannot be neglected.     

Formation Cross Sections of Zirconium-90m and Lead-207m with Fast Neutrons

Abstract

The cross sections for the formation of the metastable states of ⁹⁰Zr, (σ zr), for neutron energy of 14.8 MeV, and that of ²⁰⁷Pb, (σ pb,), for neutron energies from 4.1 to 5.2 MeV have been measured and the results are reported. The experimentally determined value of the cross section for the former is 700±94mb, while for the latter, the cross section increases from 316 + 44 mb to 788±110 mb. The cross section ratio, σpb/σZr, is also studied, which is found to vary from 1.87 to 2.18 for neutron energies from 14.8 to 18.2 MeV respectively.

The equality in the half-lives of the metastable states of ⁹⁰Zr and ²⁰⁷Pb is exploited in the present measurements, which enabled comparative evaluation of the cross sections.

Natural samples of zirconium and lead were irradiated simultaneously and the x-rays were measured off-line by a 55 cm³ Ge(Li) detector. Cyclic activation process was adopted to improve the counting statistics. The overall accuracy achieved in the measurements is about 10%.     

Neutron Irradiation Effects on Thermally Activated Process of Tensile Properties of Titanium

This paper deals with the relationship between mechanical properties and irradiation, effects for titanium irradiated to fast neutron fluxes. The neutron fluences applied are 6.9×10¹⁸, 8.6 × 10¹⁸ and 3.0 × 10¹⁹ n/cm². Tensile deformation is carried out over the temperature range of 77–about 600°K retaining the strain rate constant on one hand and changing the strain rate by a factor of about 5 and 10 on the other.

The fluence (φ) dependence of the yield stress at room temperature for an athermal component of the stress, σμ is greater than that for a thermal component σ* which does not change remarkably after irradiation. Their increments Δσ, Δμ and Δ σ* are proportional toσ ^1/3, σ^1/2σ^1/4 and, respectively.

The relationship between activation volume V* and effective shear stress τ* is investigated for both the unirradiated and irradiated specimens. In terms of the τ*/τ*₀ analysis (τ*_o is the value of τ* at T = 0°K), V* shows a tendency to decrease with increase in neutron fluence.

Irradiation defects observable by electron microscopy seem to be related to the athermal activation stress (σ_u) and those too small to be observed by electron microscopy to the thermal activation stress. The yield stress in the thermal activation can be given by Conrad's formula. The activation energy H₀ shows a constant value of about 1.8 eV irrespective of the neutron fluence applied. This value is 0.3–0.4eV higher than that for unirradiated specimens.     

Dosimetry of Fission Neutrons and Gamma-Rays from Nuclear Reactors by Paired Fricke Solutions

It is found that the heavy-water Fricke solution showed lower sensitivity to fission neutrons (G-value of 3.7) but higher sensitivity to γ-rays (G=16.3) than the light-water Fricke solution (G=5.6 for neutrons; G=15.4 for γ-rays). Using these differential G-values of the paired solutions and the basic principles of Fricke dosimeter, the following paired equations were derived for the heavy-water and the light-water solutions that were exposed to neutron-gamma mixed radiation; D_N =1,400A–1,200A′ and D_G=460A′–270A, where D_N and D_G are the absorbed doses of neutrons and γ-rays in the light-water solution, respectively, and A and A′ are absorbance increases in the light-water and the heavy-water solution, respectively. The validity of the paired equations was tested by exposure of the paired solutions to the mixed fields of nuclear reactors at Kinki University and Musashi Institute of Technology. Obtained pairs of D_N and D_G values agreed reasonably well with those measured by paired ionization chambers.     

Temperature and Concentration Dependence of Tritium Release Behavior from Solidified LiF-PbF2 Pellet Irradiated by Thermal Neutron

Thin pellets of the LiF-PbF₂ system with no porosity were irradiated by thermal neutron. Then release experiment of tritium produced in the pellets was carried out at some constant temperatures. Thus diffusion coefficients of tritium in the LiF-PbF₂ system were estimated to obtain an activation energy of diffusion, which was found to be 0.51 eV at LiF mole fraction of 0.467. It was also observed that decrease of LiF concentration in the pellet brings increase of diffusion coefficient of tritium.     

Preliminary Study on Production of Xenon-133 from Neutron-Irradiated Uranium Metal and Oxides by Oxidation

A method was developed for recovering the fission product ¹³³Xe from several kinds of reactor-irradiated U targets, including Al-U alloy, metallic U, and uranium oxides.

In order to observe the release of ¹³³Xe from U compounds at high temperatures, irradiated targets were heated at 500°~1,000°C in vacuum or under gas flow. The liberated ¹³³Xe was trapped on charcoal beds, and the release rate of ¹³³Xe from the compounds was determined by measuring the activity accumulating in the traps.

More than 90% of the ¹³³Xe was liberated from the alloy upon melting and from metal and uranium oxide upon oxidation.

The isolated Xe was purified by a system embodying cold traps and cryogenic distillation.

The final products were sealed in ampoules. They proved to possess radiochemical purity exceeding 99.9%, and less than 1 μ/ampoule (1 ml) of non-radioactive gaseous contaminants.     

Determination of Activation Cross Section by Gamma-Ray Spectrometry

A quick and easy way of determining activation cross sections by the analysis of γ-ray spectra is described. To determine cross sections by this method, it is not necessary to measure the neutron flux density, to undertake any chemical separation process, nor to use enriched target substances. To ascertain the validity of the method, the activation cross section of ¹⁰²Ru(n,γ) ¹⁰²Ru reaction was obtained, resulting in the value of 1.37±0.132 barns, which compares well with published data of 1.44±0.16 barns, determined by another method.     

Measurements of Thermal Neutron Cross Section and Resonance Integral for 237Np(n, γ)238Np Reaction

Making use of a standard neutron spectrum field with a pure Maxwellian distribution, the thermal neutron cross section for the ²³⁷Np(n, γ)²³⁸Np reaction was measured at a neutron energy of 0.0253 eV by the activation method. The result is 158±3 b, which is obtained relative to the reference value of 98.65±0.09 b for the ¹⁹⁷Au(n, γ)¹⁹⁸Au reaction. Although the data in JENDL-3 is larger by about 15% than the present value, the recently revised data in JENDL-3.2 is close to the present. The ENDF/B-V, ENDF/B-VI, JEF-2 and Mughabghab's data are also larger by 7–15%. Old measurements are larger by 7–18% than the present data.

The resonance integral for the ²³⁷Np(n, γ)²³⁸Np reaction was also measured relative to the reference value of 1,550±28 b for the ¹⁹⁷Au(n, γ)¹⁹⁸Au reaction with a 1/E standard neutron spectrum field. By defining the Cd cut-off energy as 0.5 eV for the ²³⁷Np(n, γ)²³⁸Np reaction, the present resonance integral is 652 ± 24 b, which is in good agreement with the JENDL-3, -3.2, ENDF/B-V, -VI, JEF-2 and Mughabghab's data. However, most of the old experimental data are, in general, larger by 24–38% than the present measurement.     

Surface Irradiation Damage of 316 Stainless Steel,Vanadium and Molybdenum by Helium Ion Bombardment

Molybdenum, V and 316 stainless steel were irradiated with 50~150 keV He ions at the temperatures between 413 and 1,298K for total doses ranging 1× 10²²~10ײ³ m^?2, and the characteristics of the surface damage were compared. Severe exfoliation was observed in all of these materials for the irradiation at 413±110 and 748±25K. The number of exfoliated skins was larger than that in literature, and increased nearly in proportion with the total dose. It increased in the order Mo<316SS<. When the dose was low, the amount of erosion increased rapidly with total dose, but tended to be saturated for higher doses than 3×10²² m^?2. It increased in the order Mo<V<316SS at 413±110K, while in the order 316SS<Mo<V at 748±25K. At higher temperatures than 923 K, blisters and porous surface were formed and the exfoliation of skins ceased. The amount of erosion increased with increasing incident ion energy in the energy range between 50 and 150 keV at 413±110K for a total dose of 1×10²² m^?2.     

Improved Graphite Damage Model for Predicting Property Changes of HTGR Graphites under Isothermal and Nonisothermal Irradiations

Graphite components of a High Temperature Gas-Cooled Reactor (HTGR) core will be subjected to fast neutron irradiation-induced damages causing changes in engineering properties which are dependent significantly on irradiation temperature and neutron fluence. A Graphite Damage Model (GDM) elaborated to interpret observed isothermal irradiation behavior of various polycrystalline graphites has been improved in order to apply it to nonisothermal irradiations. The paper outlines the physical and mathematical formulations of the improved GDM together with some comparisons between predictions and measurements. On the basis of nonlinear fittings of the model functions to measured data, a total of 28 global modeling parameters have been determined successfully. Furthermore, a recurrence formula has been devised to permit a nonisothermal irradiation behavior to be predicted in terms of the new GDM. This has been proved by comparison of the predicted changes in linear dimension and thermal conductivity with the measured ones of some graphite materials irradiated at constant and changing temperatures.     

Snoek Peak of Neutron-Irradiated Fe-3.48% Ni Alloy

The Snoek peak of a neutron-irradiated Fe-3.48% Ni alloy was measured to determine the resolution of interstitial impurity atoms.

The internal friction was measured after isothermal annealing on specimens irradiated at 75° ± 10°C to a dose of ～3.6 × 10¹⁹ nvt.

The change of the Snoek peak height, which corresponds to the amount of nitrogen in solution, was analyzed upon assuming the applicability of the rate equation.

As a result of calculation, the order of the reaction was found to be about 2.6, and the activation energy for resolution of nitrogen atoms about 2.9 eV.     

Separation of Pu-238 and Pu-242 from Irradiated Am-241

Failure analysis was made on samples taken from type 304 stainless steel piping systems (core spray, unloading and feed water pipes) that had cracked in service. In the core spray pipe, large cracks including one penetrating through the wall were found in the upper half of the pipe wall, within the heat-affected zone of the weld joint between the pipe and the nozzle safe end of the reactor pressure vessel. These cracks were of intergranular type, also small transgranular cracks were found beyond the heat-affected zone. Strong correlation was established between the intergranular cracking and severe sensitization in the heat-affected zone.

In the unloading pipe, the region close to the weld joint contained cracks similar to those in the core spray pipe. The feed water pipe, in contrast to the foregoing cases, contained numerous shallow transgranular cracks both within and outside the heat-affected zone.     

Effect of Pre-Hydriding on Post-Irradiation Ring-Tensile Properties of Zircaloy-2 Cladding Tubes

Abstract

Zircaloy-2 tubes were hydrided up to a nominal content of 200 ppm and irradiated as fuel claddings in HBWR. Post-irradiation ring-tensile testing revealed that hydrogen enhances the irradiation-induced decrease of elongation and wall thickness reduction at room temperature. On the other hand, no effect of hydrogen was observed on ultimate tensile strength. With testings at 300°C, the effect was negligible on elongation too. From the evaluation of the test results including metallographic observation of ring specimens after fracture, it was concluded that segregation of hydrides due to thermal diffusion of hydrogen during irradiation was at least a part responsible to the above effect of hydrogen enhancing embrittlement.     

Analysis of Neutron and Secondary Gamma-Ray Transport in Air-Over-Ground Geometry by PALLAS-2DCY-FC Code

Comparisons are made between PALLAS calculated with meaured neutron and γ-ray doses above the ground in an air-ground medium for HENRE accelerator and BREN reactor, yielding good agreements except for a 8.23-m height of HENRE source, in which the calculation overestimates the neutron dose by a factor of 1.5 due to the use of a rough angular quadrature set. Disregard of the ground results in a decrease by a factor of 2 in both neutron and γ-ray doses compared with those for the presence of the ground, while for an infinite-air medium both these doses increase with distances from the source, which indicates that the ground should not be. ignored in the neutron and secondary p-ray transport calculation. For assumed neutron skyshine calculations disregard of the ground results in a decrease in the neutron dose by a factor of 1.5 and 1.7 and also in the secondary γ-ray dose by a factor of 1.35 and 3~5 respectively for a 14-MeV source and a fission source. In addition to the importance of inclusion of the ground in neutron skyshine calculations, an additional essential factor is the secondary γ-ray production due to neutron inelastic scattering interaction with nitrogen for the 14-MeV source, or the one due to neutron capture interactions in both ground and air for the fission source.