今日的尤卡山计划和多重屏障系统的新概念

主要论述了美国尤卡山项目的最新进展和比利时工程屏障研究的新进展。在尤卡山项目新进展方面介绍了尤卡山项目修改的近期计划,以及两个环境评价补充报告和两位民主党参议员参加2008年美国总统竞选时对尤卡山项目的态度。还介绍了比利时为加强高放废物地质处置的安全性而开发的多重屏障的新概念。     

Failure model and Monte Carlo simulations for titanium (grade-7) drip shields under Yucca Mountain repository conditions

A failure model was developed for the titanium alloy drip shield proposed for the Yucca Mountain nuclear waste repository. The degradation modes considered are hydrogen-induced cracking (HIC) and general aqueous corrosion, processes which are inextricably linked. Failure by HIC is controlled by the environment, the corrosion rate, the material properties, and the hydrogen absorption efficiency which is assumed to decrease parabolically with time. This model includes both oxygen and water reduction coupled to corrosion, and allows for the release of the absorbed hydrogen as the alloy containing the hydrogen converts to oxide. Monte Carlo simulations were employed to predict drip shield lifetimes, and to investigate the effects of the hydrogen absorption efficiency, the critical HIC concentration, the corrosion rate, and the fraction of corrosion supported by water reduction, on the susceptibility of the material to HIC.     

美国尤卡山高放废物处置库项目费用削减的原因分析

介绍了美国尤卡山高放废物处置库项目2009年和2010年费用遭到削减这一事件的来龙去脉,从政治、情感、技术、经济和资源再利用几个方面分析了尤卡山项目经费削减的原因,指出美国参议院多数党领袖里德的激烈反对以及政治考虑是导致费用削减的主要原因。同时介绍了一系列反对削减尤卡山项目经费的观点。     

Understanding long-term corrosion of Alloy 22 container in the potential Yucca Mountain repository for high-level nuclear waste disposal

Alloy 22 (Ni–22Cr–13Mo–3W–4Fe) is the candidate material for the waste package outer container in a potential geologic repository for high-level nuclear waste disposal at Yucca Mountain, Nevada. This alloy exhibits very low corrosion rates in the absence of environmental conditions promoting crevice corrosion. However, there are uncertainties regarding Alloy 22’s corrosion performance when general corrosion rates and susceptibility to crevice corrosion are extrapolated to a geological time period (e.g. 10⁵ years). This paper presents an analysis of available literature information relevant to the long-term extrapolation of general corrosion processes and the crevice corrosion behavior of Alloy 22, under potential repository environments. For assessment of general corrosion rates, potential degradation processes causing the loss of the long-term persistence of passive film formed are considered. For crevice corrosion, induction time, and the extent of susceptibility and opening area, are considered. Disclaimer: The US Nuclear Regulatory Commission (NRC) staff views expressed herein are preliminary and do not constitute a final judgment or determination of the matters addressed nor of the acceptability of a license application for a geologic repository at Yucca Mountain. The paper describes work performed by the Center for Nuclear Waste Regulatory Analyses (CNWRA) for NRC under Contract Number NRC-02-02-012. The activities reported here were performed by CNWRA on behalf of the NRC office of Nuclear Material Safety and Safeguards, Division of High Level Waste Repository Safety. This paper is an independent product of the CNWRA and does not necessarily reflect the view or regulatory position of the NRC.     

Analysis of thermo-hydro-mechanical process in the engineered barrier system of a high-level waste repository

This paper presents a simulation of the KAERI's engineering scale experiment, KENTEX which is to study the thermal, hydrological and mechanical (THM) processes occurring in the engineered barrier system of a high-level waste repository. The simulation was performed by using the computer code, TOUGH2, which analyzes the multi-dimensional fluid and heat flows of multiphase, multicomponent fluid mixture in unsaturated medium. The detailed geometry of KENTEX was incorporated into the model, and the laboratory experiments were carried out to determine the thermal, hydrological and mechanical properties of bentonite, which were used as input parameters for the simulation. The calculated results on the temperature, water content, and total pressure distribution throughout the bentonite buffer were compared with the experimental ones.     

Deterministic Assessment for Environmental Impact of Yucca Mountain Repository Measured by Radiotoxicity

Environmental impact of the Yucca Mountain Repository (YMR) has been quantitatively and deterministically evaluated in terms of the radiotoxicity of transuranic (TRU) and fission-product radionuclides existing in the environment after released from failed packages. Inventory abstraction has been made based on the data published in Final Environmental Impact Statement by US Department of Energy (DOE). Mathematical model and computation code have been developed based on analytical solutions. Environmental impact from the commercial spent nuclear fuel (CSNF) packages is about 90% of the total impact including the contribution from defense waste (DW) packages. Impacts due to isotopes of Cm, Am, Pu and Np, and their decay daughters are dominant, compared with those from fission-product nuclides. Numerical results show that reduction of the TRU nuclides by a factor of 100 makes the impact from CSNF smaller than that from DW.     

Second generation waste package design and storage concept for the Yucca Mountain Repository

The reference waste package design and operating mode to be used in the Yucca Mountain Repository is reviewed. An alternate (second generation) operating concept and waste package design is proposed to reduce the risk of localized corrosion of waste packages and to reduce repository costs. The second generation waste package design and storage concept is proposed for implementation after the initial licensing and operation of the reference repository design. Implementation of the second generation concept at Yucca Mountain would follow regulatory processes analogous to those used successfully to extend the design life and uprate the power of commercial light water nuclear reactors in the United States. The second generation concept utilizes the benefits of hot dry storage to minimize the potential for localized corrosion of the waste package by liquid electrolytes. The second generation concept permits major reductions in repository costs by increasing the number of fuel assemblies stored in each waste package, by eliminating the need for titanium drip shields and by fabricating the outer container from corrosion resistant low alloy carbon steel.     

Calculation of maximum release rates in alternative design changes in the thickness of the buffer for the engineered barrier system (EBS) in deep repository by using Amber code

The multibarrier concept forms the basis for geological disposal concepts in most countries and the guideline states that research and development should aim to demonstrate the feasibility of constructing an engineered barrier system (EBS) which is appropriate for the range of relevant geological conditions. The multibarrier system (EBS) and its functions consisting of the glass waste form, overpack and buffer material was located in a sufficiently stable geological environment. When an overpack comes into contact with groundwater it will start to corrode. The wall thickness will then gradually reduce, and the overpack will eventually fail mechanically when its structural strength can no longer support the stress imposed by the surrounding environment. The requirements that influence the thickness of buffer include nuclide migration retardation and heat conductivity, as well as stress buffering capability, self-sealing ability and workability. The migration retardation function is assumed to be the most important of all these requirements with respect to setting the appropriate thickness of buffer. A consideration of these effects and relationship between buffer thicknesses has determined that a reasonable thickness for the buffer is between 400 mm and 700 mm [AECL, H12]. Therefore, the design thickness of buffer material can range from 0.4 m to 0.7 m to account for manufacturing and stress buffering. In this alternative design case, the thickness of the buffer material is set to 0.4 m, 0.5 m, 0.6 m and 0.7 m. The nuclide migration properties of the buffer material are assumed to be the same (PNC, Development and Management of the Technical Knowledge Base for the Geological Disposal of HLW, Supporting Report 2: “Repository” Engineering Technology). The results of calculation are presented that some nuclides such as Se-79, Tc-99, Pd-107, Th-233, U-236, Pb-210, Ra-226 and Np-237 virtually unchanged in case the maximum release rate from EBS corresponding to change thickness of buffer material. Some nuclides such as Cs-135, Nb-94, Nb-93 m, Zr-93, Sn-126, Th-230, Ph-240, Pu-242, U-233, Ac-227, Pa-231 and Th-229 are very little greater for 40 cm, 50 cm and 60 cm in the maximum release rate compared with 70 cm. Maximum release of nuclides U-235, U-234 and U-238 increases in case of 50 cm and 60 cm thickness of buffer and in case 40 cm are the same as 70 cm thickness because the amount of their parents in case 40 cm will decrease before decay and in case 70 cm amount of these nuclides will decrease due to decayed to other nuclides before release from the buffer, then maximum decay happened in case 50 cm. The maximum release rates of short-lived nuclides such as Cm-245, Am-243, Cm-245, Am-241, Pu-241 and Pu-239 increase significantly due to less decay occurring during the reduced buffer transit time.     

Design and corrosion study of a closed spallation target module of an accelerator-driven system (ADS)

At the Forschungszentrum Karlsruhe (FZK) the characteristics of an accelerator-driven subcritical reactor system (ADS) are critically evaluated, mainly with respect to the potential of transmutation of minor actinides and long-lived fission products, to the feasibility and to safety aspects. The work is concentrating on system design, neutronics, thermalhydraulics, safety, materials and corrosion. This article describes the FZK approach to design a closed 4 MW(th) spallation target module with a solid beam window and eutectic lead–bismuth (Pb–Bi) as spallation material and cooling fluid, which is going to be implemented in the FZK three-beam concept of an ADS. This multi-beam concept shows significant improvements towards single-beam concepts from the literature with respect to power distribution in the subcritical blanket and thermal loads of heat removal from the beam window and the spallation region. For some selected martensitic and austenitic steels, corrosion tests in static lead are performed to examine their suitability as structural or window materials. Alloying aluminum into the surface layer by high-power electron beam treatment, corrosion can be reduced to nearly zero. One prerequisite to minimize corrosion is a proper oxygen control system (OCS) via the gas-phase to set the oxygen concentration in the liquid Pb–Bi. The dynamic behaviour of this oxygen control system is described. Finally, the KArlsruhe Lead LAboratory (KALLA) is introduced, the objectives of which are technological, thermal-hydraulic and corrosion investigations into the beam window, the spallation target module and the primary system of an ADS.     

A mathematical model for crevice corrosion under porous deposits

A model has been developed to investigate the corrosion of steels in a thin, narrow crevice formed between the metal surface and an oxygen-permeable, porous deposit. A thin electrolyte layer exists within the deposit, due to geochemical fluids dripping onto a deposit-covered surface or due to the adsorption of moisture by a hygroscopic deposit. Mass transfer by diffusion and ion migration is considered in both the electrolyte films inside and outside of the crevice. The main reactions considered are the anodic dissolution of the alloy substrate, hydrolysis of the alloying element cations, dissociation of water, and the cathodic reduction of oxygen, hydrogen ion, and water. Special attention has been given to the role of parameters connected with the porous layer (porosity, tortuosity, and the layer thickness) on the rate of crevice corrosion. It is shown that the cavity acts as an ‘electrochemical amplifier’ from the point of view of the concentration of aggressive anions that leads to increasing of corrosion rate and to a higher probability of pit nucleation within the crevice.     

On-line monitoring system development for single-phase flow accelerated corrosion

Aged nuclear piping has been reported to undergo corrosion-induced accelerated failures, often without giving signatures to current inspection campaigns. Therefore, we need diverse sensors which can cover a wide area in an on-line application. We suggest an integrated approach to monitor the flow accelerated corrosion (FAC) susceptible piping. Since FAC is a combined phenomenon, we need to monitor as many parameters as possible and that cover wide area, since we do not know where the FAC occurs. For this purpose, we introduce the wearing rate model which focuses on the electrochemical parameters. Using this model, we can predict the wearing rate and then compare testing results. Through analysis we identified feasibility and then developed electrochemical sensors for high temperature application; we also introduced a mechanical monitoring system which is still under development. To support the validation of the monitored results, we adopted high temperature ultrasonic transducer (UT), which shows good resolution in the testing environment. As such, all the monitored results can be compared in terms of thickness. Our validation tests demonstrated the feasibility of sensors. To support direct thickness measurement for a wide-area, the direct current potential drop (DCPD) method will be researched to integrate into the developed framework.     

Radiation-induced ‘long-cell’ (macro-cell) corrosion in light water reactors

In previous papers, the author has established various ‘long-cell’ (akin to ‘macro-cell’) corrosion configurations that exist in nuclear power plants. Among these, the radiation-induced corrosion cell is an important mechanism since it plays a major role in the corrosion problems found in primary water of the nuclear power plants. There are numerous experimental evidences indicating a potential difference induced by radiation, however, the exact mechanism of such phenomena has not been clarified. The author investigated the basic mechanism by combining radiation chemistry, electrochemistry and corrosion science to confirm the existence of radiation-induced ‘long-cell’ action.By performing a competition kinetic study, , reacting mainly with stable molecules are found to be responsible for inducing a large portion of the potential difference both in the PWR and BWR water chemistry environments. The hydrated electrons react at a cathodic half-cell thereby inducing reductive reactions in the mixed cell configuration. This method reproduces the reported and experimentally observed redox potential variation to a certain extent (observed in the INCA Test Loop in Sweden and the NRI-Rez BWR-2 Loop in Czech Republic). The author believes the results support the assumed corrosion mechanism although details are still debatable.     

Hybrid soliton nuclear reactors: A model and simulation (encapsulated long living accelerator driven system)

We intend to explore the potential of Hybrid Soliton Reactors (Réacteur Hybride à Soliton, RHYS) for producing energy. In our case an encapsulated long living fission reactor is driven by a proton accelerator, who produces neutrons on a target. In a first part we give the mathematical approach of such a sub-critical reactor, as an extension of the “Soliton Reactor” which was recently proposed by different authors, Edward Teller, L.P. Feoktistov, and others (H. Sekimoto under the name “Candle reactor”). In a second part we give results of simulations and explore the possibilities to control such a system.     

A model for uniform Zircaloy clad corrosion in pressurized water reactors

A phenomenological water-side corrosion model for Zircaloy fuel cladding for pressurized water reactors (PWRs) is considered. The model acounts for the breakaway transition in the Zircaloy oxidation rate that takes place in an isothermal condition and the changes that occur during reactor operation, i.e. the dependence of oxide growth on fast neutron flux and cladding oxide layer thickness. Closed-form analytical solutions of the oxidation kinetics equations are obtained. The corrosion kinetics model is coupled to PWR thermal and hydraulic models which assume a subchannel that is either a closed single channel or a multichannel which accounts for coolant cross-flow and coolant enthalpy mixing. Both single-phase forced convection and subcooled nucleate boiling are accounted for in the thermal-hydraulic models. The model calculates the coolant temperature at the axial midplane of each axial segment of the fuel rod. When an oxide layer is present, the temperature at the metal-oxide interface is determined. This temperature in turn is used to determine the oxide growth via the Arrhenius temperature dependence of the Zircaloy oxidation rate. The predictions of the model have been compared with the measured cladding oxide data obtained in PWRs. The data for a given rod were obtained at various burn-ups (at the end of reactor cycles) and various axial positions of the rod. Our evaluations show that the model predicts the measured data satisfactorily; however, the deviations are discussed. The model has been used to study the effect of core loading patterns on cladding oxide growth. Our analyses show that core nuclear design is an important factor for water-side corrosion of fuel rods.     

The Thermal-Hydraulic model for the pebble bed modular reactor (PBMR) plant operator training simulator system

This paper provides a discussion of the model development status and verification efforts for the Reactor Core Thermal-Hydraulic model developed for the full-scope plant Operator Training Simulator System of the Pebble Bed Modular Reactor (PBMR). Due to the First of a Kind Engineering nature and lack of reference plant data, model verification has mainly been focused on benchmarking the model configurations against test cases performed by PBMR design analysis codes, i.e. TINTE, VSOP and FLOWNEX.As a first step, due to the symmetrical physical nature of the PBMR core, a two-dimensional (2D) model configuration in radial and axial directions (axial-symmetry) was developed. The design was subsequently extended to a three-dimensional (3D) configuration. Through the use of cross-flow and cross-conduction links, three nearly identical 2D configurations were glued together to form this 3D model configuration. To date, the 3D configuration represents the most comprehensive model to simulate the PBMR core thermo-hydraulics. This paper concludes with the verification of thermodynamic and heat-transfer properties of two steady state (100% and 40% power) conditions between the 3D Reactor Core Thermal-Hydraulic model and the available FLOWNEX and TINTE design code analysis. The transient operations between these two power levels are also discussed.     

纳米及超细晶结构锆金属腐蚀速率计算模型

本文在Wagner理论和固体电子理论的基础上,将电导率与电子的平均自由程及晶粒尺寸的关系联系起来,建立了纳米及超细晶结构金属腐蚀速率一晶粒尺寸关系模型,并结合锆金属的性质,模拟计算了不同温度下晶粒尺寸对锆合金腐蚀速率常数的影响,结果表明,与普通品粒相比,纳米晶粒尺寸下锆Zr的腐蚀速率常数和腐蚀增重,远低于普通晶粒纯Zr的腐蚀速率常数和腐蚀增重,同时,随着纳米晶粒尺寸的减小Zr的腐蚀速率常数也降低,显示晶粒纳米化处理可以改善锆金属的腐蚀性能.     

Pulsed laser deposition of corrosion protective Yttrium Oxide (Y2O3) coating

Among ceramics and oxides, Yttria (Y₂O₃) films have been widely investigated as potential corrosion protective coatings largely on account of their wear resistant and non-wettable properties. Results presented here describe successful use of pulsed laser deposition (PLD) technique for deposition of thin film Yttria coating on stainless steel substrates. Deposited Yttria coatings have been characterized in terms of their microstructure, crystalline phase and hardness using Scanning Electron Microscope (SEM), X-ray diffraction (XRD) and scratch test techniques, respectively. Characterization tests of these coatings of thickness up to 50 μm have shown strong bonding with substrate surface and a high degree of homogeneity and compaction. Resistance of these PLD based Yttria coatings to molten uranium have also been studied via Differential Thermal Analysis (DTA). Our results on DTA tests evaluating compatibility of the Yttria coatings with molten uranium have established the excellent corrosion resistance property of such Y₂O₃ coatings when exposed to molten uranium.     

Design features of a cable–bollard vehicle barrier system

The design features of the stand-alone cable–bollard vehicle barrier system (Cable-Bollard VBS) developed for the Vermont Yankee Nuclear Power Plant (VY) to meet the design goals of the recent 10 CFR Part 73 rule changes are discussed. The design is based on the application of fundamental engineering principles to a dynamic system, recognizing that vehicle impact on a cable system is fundamentally different from vehicle impact on a bollard or other hard barrier. As such, rigorous attention is paid to cable anchor design and performance.