Effects of α-radiation on a direct disposal system for spent nuclear fuel – (2) review of research into safety assessments of direct disposal of spent nuclear fuel in Europe and North America

The Japanese geological disposal programme has started researching disposal of spent nuclear fuel (SF) in deep geological strata (hereafter “direct disposal of SF”) as an alternative management option other than reprocessing followed by vitrification and deep geological disposal of high-level radioactive waste (HLW). In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include: increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with radiation degradation of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate and the solubility of SF. Therefore, the influences of radiation, which are not expected to be significant in the case of geological disposal of vitrified waste, must be considered in safety assessments for direct disposal of SF. Focusing especially on the effects of α-radiation in safety assessment, this study has reviewed safety assessments in countries other than Japan that are planning direct disposal of SF. The review has identified issues relevant to safety assessment for the direct disposal of SF in Japan.     

Neutronic model of a mirror based fusion–fission hybrid for the incineration of the transuranic elements from spent nuclear fuel and energy amplification

The Georgia Institute of Technology has developed several design concepts of tokamak based fusion–fission hybrids for the incineration of the transuranic elements of spent nuclear fuel from Light-Water-Reactors. The present paper presents a model of a mirror hybrid. Concerning its main operation parameters it is in several aspects analogous to the first tokamak based version of a “fusion transmutation of waste reactor”. It was designed for a criticality k_eff ≤ 0.95 in normal operation state. Results of neutron transport calculations carried out with the MCNP5 code and with the JEFF-3.1 nuclear data library show that the hybrid generates a fission power of 3 GW_th requiring a fusion power between 35 and 75 MW, has a tritium breeding ratio per cycle of TBR_cycle = 1.9 and a first wall lifetime of 12–16 cycles of 311 effective full power days. Its total energy amplification factor was roughly estimated at 2.1. Special calculations showed that the blanket remains in a deep subcritical state in case of accidents causing partial or total voiding of the lead–bismuth eutectic coolant. Aiming at the reduction of the required fusion power, a near-term hybrid option was identified which is operated at higher criticality k_eff ≤ 0.97 and produces less fission power of 1.5 GW_th. Its main performance parameters turn out substantially better.     

Applications of (n,p) and (n, α) reactions and a backscattering technique to fusion reactor materials,archeometry, and nuclear spectroscopy

Depth profiles of He, Li and B are determined by ³He(n, p)T, ⁶Li(n, α)T and ¹⁰B(n, α)⁷Li reactions with thermal neutrons at the high flux reactor of the ILL, Grenoble. The behaviour of Li in Be is examined with respect to future fusion reactors. Range profiles of 70–300 keV Li⁺ are measured and found to agree with theory based on Lindhard-Scharff electronic stopping and Molière potential. Li becomes mobile in Be above 100°C. Further, B and Li distributions in glaze of ancient pottery are examined for studying ancient production techniques. It is found that all examined samples (of Islamic, Thai and North American provenience) show Li and B concentrations which are enriched relative to the original material. Li is mostly depleted in a surface layer of 0.1–1.6 μm half-width due to various burning conditions.In experimental nuclear physics, gas cells are now often replaced by thin foils with implanted gas. In many cases the knowledge of the concentration profile is required, and is presently evaluated for the case of ³He in Ni and Au with the (n, p) reaction. This is compared to results obtained by a special Rutherford backscattering technique yielding good agreement.     

Methodology and application of the nuclear resonance reaction 16O(α, α)16O for the profiling of titanium oxide

A method is presented for the application of the ¹⁶O(α,α)¹⁶O resonance at 3.045 MeV to the depth profiling of oxide films. The resonance yield is translated into oxygen to metal stoichiometry as a function of the depth probed. The method is applied to anodic titanium oxide films grown in the absence and presence of tungsten anions. The oxygen depth profile is then utilized to obtain the profile of the dopants (W) incorporated in the oxide. A depth resolution of 6 μg/cm is achieved. The incorporation of tungsten anions of the order of 2 at.%, evaluated from the RBS yields, precludes the establishment of a constant oxygen to metal stoichiometry in the doped film. The tungsten atomic fractions have an estimated error of 5%.     

Experience in operating and upgrading the No. 5 unit of the Novovoronezh nuclear power plant – practical base for developing a reliable source of nuclear energy

The No. 5 unit of the Novovoronezh nuclear power plant, starting commercial operations on September 26, 1980, is the first power-generating unit with a 1000 MW VVER in our country. The assimilation of its power gave invaluable experience to designers, builders, and equipment manufacturers; this experience was taken into account in the design solutions for next-generation power-generating units. A large volume of work on increasing the efficiency, reliability, and safety was performed over a 30-year service life. At present, the power-generating unit has been shut down for a major overhaul for upgrading according a program for extending the service life by 25–30 years.     

The Effects of A1, Cr on the Oxidation Behavior of a V-4Ti Alloy

Oxidation experiment was performed in air at elevated temperature for three kinds of vanadium alloys. The features of the oxides and the role of the alloying elements were analyzed. All specimens exhibited a parabolic kinetic behavior of weight gain with the exposure time. The alloys can‘t be used in air at the temperature over 700℃, below which, V4Ti3A1 showed a much lower oxidation rate than the V4Cr4Ti alloy. It was found that A1 in the alloy was segregated to the specimen surface in the process, and formed into Al2O3 on the surface, hence decreasing the formation of V2O5. The oxides on the surface were nucleated in a small number density and grew to a large size, giving more protection to the matrix alloy.     

Can redox sensitive radionuclides be immobilized on the surface of spent nuclear fuel? - A model study on the reduction of Se(IV)aq on Pd-doped UO2 under H2 atmosphere

Spent nuclear fuel contains noble metal particles composed of fission products (Pd, Mo, Ru, Tc, Rh and Te, often referred to as ε-particles). Studies have shown that these particles play a major role in catalyzing oxidative dissolution as well as H₂ reduction of the oxidized UO₂ fuel matrix, depending on the conditions. Thus it is possible that these particles also could have a major impact on the state of other redox sensitive radionuclides (such as the long lived fission product ⁷⁹Se) present in spent nuclear fuel. In this study, Pd-doped UO₂ pellets are used to simulate noble metal particles inclusions in spent nuclear fuel and the effect on dissolved selenium in the form of selenite (250 μM selenite) in simulated ground water solution (10 mM NaCl, 10 mM NaHCO₃) at 1 and 10 bar hydrogen pressure. The selenite was found to be reduced to elemental Se, forming colloidal particles. At hydrogen pressures of 10 bar, the rate of selenite reduction was found to be linearly correlated to the fraction of Pd in the UO₂ pellets. No selenium was detected on the surface of the pellets. For the lowest Pd loading (0.1% Pd) the selenite reduction does not appear to proceed to completion indicating that the surface becomes less active.     

Validation of a CATHENA fuel channel model for the post blowdown analysis of the high temperature thermal–chemical experiment CS28-1, I – Steady state

To form a licensing basis for the new methodology of the fuel channel safety analysis code system for CANDU-6, a CATHENA model for the post-blowdown fuel channel analysis for a Large Break LOCA has been developed, and tested for the steady state of a high temperature thermal–chemical experiment CS28-1. As the major concerns of the post-blowdown fuel channel analysis of the current CANDU-6 design are how much of the decay heat can be discharged to the moderator via a radiation and a convective heat transfer at the expected accident conditions, and how much zirconium sheath would be oxidized to generate H₂ at how high a fuel temperature, this study has focused on understanding these phenomena, their interrelations, and a way to maintain a good accuracy in the prediction of the fuel and the pressure tube temperatures without losing the important physics of the involved phenomena throughout the post-blowdown phase of a LBLOCA. For a better prediction, those factors that may significantly contribute to the prediction accuracy of the steady state of the test bundles were sought.     

Burnable absorber-integrated Guide Thimble (BigT) – I: design concepts and neutronic characterization on the fuel assembly benchmarks

This paper presents the conceptual designs of a new burnable absorber (BA) for the pressurized water reactor (PWR), which is named ‘Burnable absorber-integrated Guide Thimble’ (BigT). The BigT integrates BA materials into standard guide thimble in a PWR fuel assembly. Neutronic sensitivities and practical design considerations of the BigT concept are points of highlight in the first half of the paper. Specifically, the BigT concepts are characterized in view of its BA material and spatial self-shielding variations. In addition, the BigT replaceability requirement, bottom-end design specifications and thermal–hydraulic considerations are also deliberated. Meanwhile, much of the second half of the paper is devoted to demonstrate practical viability of the BigT absorbers via comparative evaluations against the conventional BA technologies in representative 17×17 and 16×16 fuel assembly lattices. For the 17×17 lattice evaluations, all three BigT variants are benchmarked against Westinghouse's existing BA technologies, while in the 16×16 assembly analyses, the BigT designs are compared against traditional integral gadolinia-urania rod design. All analyses clearly show that the BigT absorbers perform as well as the commercial BA technologies in terms of reactivity and power peaking management. In addition, it has been shown that sufficiently high control rod worth can be obtained with the BigT absorbers in place. All neutronic simulations were completed using the Monte Carlo Serpent code with ENDF/B-VII.0 library.     

Use of the (α, n) reaction for the quantitative determination of the beryllium,boron and fluorine content of concentration products

A rapid method for the quantitative analysis of the beryllium, boron and fluorine contents in concentration products, based on the use of the (α,n) reaction, is described. Po²¹⁰ with an activity of 250 μ Cwas used as the α-radiation source. The number of neutrons in the concentration products knocked out by α-particles is recorded by a neutron counter and serves as a measure of the beryllium, boron, and fluorine content in the product analyzed. The method makes it possible to check the concentration products of beryllium, fluorite and boron ores, and also rich ores containing fluorine and boron, with an accuracy up to 1.5–2%. The content of the element checked can be determined in the presence of a constant amount of admixtures of other elements with a high(α,n) reaction yield.     

The Design,Manufacture and Testing of a Container Module to Satisfy the IAEA Regulations for the Safe Transport of Radioactive Material – 1985 Edition (as Amended 1990), for the Transport of an Advanced Gas Cooled Reactor (AGR) Gas Circulator

Abstract

Heysham 1 and Hartlepool Nuclear Power Stations share a common design of gas circulator and a requirement was identified to transport a spare circulator between the Stations. The circulators are 6 m long, 2 m diameter and weigh approximately 38 tonnes. The circulator becomes contaminated in use and is classified as suitable for transport as SCO-II. It can therefore be transported within a Type 2 Industrial Package (IP-2), in accordance with the IAEA Transport Regulations. An appropriate package, the Gas Circulator Transportation Module, based on ISO container standards, was designed, manufactured and tested, and a Safety Case prepared addressing the radiological hazards associated with the transport of a gas circulator and demonstrating regulatory compliance. Following the issue of a Certificate of Regulatory Compliance, the container module was used· to transport a gas circulator from Heysham 1 to Hartlepool, paving the way for similar purpose-designed modules for the transport of large components.     

Evaluation of Activation Cross Sections for (n,α) and (n,n'α) Reactions on ~(63, 65, Nat)Cu

Introduction Copper is a very important structure material in nuclear fusion engineering. The neutron activation cross section are very useful in fusion research and other applications such as radiation safety, environmental, material damage and neutron dosimetry. More efforts are required to identify and resolve the differences and discrepancies in the existing activation cross sections from different laboratories. The natural copper consists of two stable isotopes, i.e. ~(63)Cu, ~(65)Cu. Their abundaces and threshold energies are given in Table 1.     

Evaluation of Activation Cross Sections for (n,p) and (n,n′p) Reactions on ~(63,65,Nat)Cu

Introduction Copper is a very important structure material in nuclear fusion engineering. The neutron activation cross section are very useful in fusion research and other applications such as radiation safety, environmental, material damage and neutron dosimetry. More efforts are required to identify and resolve the differences and     

Evaluation of the (n,α) Cross Sections for ~(58, 60～62, 64)Ni and ~(Nat)Ni

The cross section of the (n,α) reaction is very important for fusion reactor for monitoring neutron field in the context of radiation induced material damage, radiation safety, neutron dosimetry, etc. The natural nickel consists of five stable isotopes, i. e. ~(58)Ni,~(60～62,64)Ni. Their abundances and threshold energies are given in Table 1.     

Magnetovolume and chemical bonding effects of Sn atom in the γ′-(Fe_(1-x)Sn_x)_4N compounds

1 INTRODUCTIONIt is known that solid solution is fOrmed by introducing some additive atoms into abasic phase. Cornparing with the origina1 material, the new mateIial fOImed b} introduc-ing interstitial atoms or substitutional atoms has been changed in the unit cel1 volurneand the charge distribution. As the structure type retains the same befOre and after tlleintroduction, the influence of introduced atom can be divided into two parts.\" One ismagnetovolame efFect (MVE) which comes fro…