Current research on environmentally assisted cracking in light water reactor environments

The effect of dissolved oxygen level on fatigue life of austenitic stainless steels is discussed and the results of a detailed study of the effect of the environment on the growth of cracks during fatigue initiation are presented. Initial test results are given for specimens irradiated in the Halden reactor. Impurities introduced by shielded metal arc welding that may affect susceptibility to stress corrosion cracking are described. Results of calculations of residual stresses in core shroud weldments are summarized. Crack growth rates of high-nickel alloys under cyclic loading with R ratios from 0.2 to 0.95 in high-purity water that contains <5 and 300 ppb dissolved oxygen at 240, 289, and 320°C, are summarized.     

聚变堆用典型金属工程材料辐照损伤机理

聚变堆中极端辐照环境下,核工程材料的安全与可靠性对保障核能产业可持续发展具重大意义。采用蒙特卡罗程序包Geant4建立了聚变堆辐照环境下的材料损伤模型,从材料的位移损伤率和杂质沉积等方面研究了CLAM钢、F82H钢、?-Fe三种聚变堆用典型金属工程材料分别在中子、质子、重离子轰击下的辐照损伤机理。研究表明,中子对材料的辐照损伤主要为位移损伤;质子和重离子对材料造成的位移损伤呈Bragg峰曲线分布,且损伤区域与粒子射程均集中在材料表层,其中14.67 Me V质子射程为512?m,0.82 Me V 3He离子射程仅为2.1?m。系统分析了聚变堆用典型金属工程材料的损伤形成机理,为进一步研究材料受辐照后宏观性能与微观结构变化提供了理论依据。     

Development of constitutive equations for nuclear reactor core materials

A set of strain rate dependent constitutive equations has been described which is capable of predicting deformation behavior of anisotropic metals under complex loading conditions with or without the presence of a neutron flux. The important feature of the constitutive equations is that they describe history dependent plastic deformation behavior of anisotropic metals under three-dimensional stress states. Since the analytical model accounts for the effect of prior deformation history at all times, it is capable of handling consecutive or simultaneous loading histories, such as post-irradiation loading, in-pile loading, etc. It is demonstrated that the general form of the constitutive relations is consistent with experimental observations made for Zircaloys under both unirradiated and irradiated conditions.     

A reactor system for radiation research and materials testing

Translated from Atomnaya Énergiya, Vol. 68, No. 6, pp. 438–441, June, 1990.     

Efficient code system of core characteristics for small cylindrical thermal reactor

A new code system for the overall neutronic calculation of a thermal reactor by a simple and effective way is presented. The code covers microscopic library compilation, macroscopic constant generation, cell calculations by multi-group treatment for neutron transport equation and core calculations over three zones for fuel and one zone for moderator. The Dancoff correction factor required in the interpolation of the self-shielding factors of resonance nuclides is automatically calculated by the installed collision probability routines. The burn-up calculation and Garrison and Ross model of fission product have been included. Also the effect of control rod on the reactivity of the reactor with special treatment for the control rod based on the homogenization technique has been included. Making a comparison with SRAC95 code system has checked the adopted code.     

Characteristics of fuel melting and relocation in metallic-fueled fast reactor core and its feasibility for eliminating recriticality

The paper gives a concept for eliminating recriticality so that a fast reactor core can have high potential to terminate severe accidents and prevent their progression into further disruption leading to a large-scale inter-subassembly molten configuration. The basic idea to eliminate recriticality events is to remove a certain amount of fuel materials out of the core in order to keep the core subcritical. This concept is called as Controlled Material Relocation (CMR). Based on the concept, we propose an example of CMR devising consisting of annular fuel pins for a metallic-fueled fast reactor core. By this devising, we can define the fraction of core fuels that should be preferentially removed out of the core to eliminate the recriticality potential. We have analyzed melting and relocation behavior of metallic fuel pin, which has annular fuel meat, under accidental condition leading to core degradation. We have also examined the possibility of eliminating neutronic recriticality for a metallic-fueled fast reactor, where the annular fuel pins are partially embedded in the core for realizing the CMR concept.     

Apparatus for the investigation of cracking of materials in thin-walled elements in the core of a water-moderated,water-cooled power reactor (VVÉR)

Translated from Atomnaya Énergiya, Vol. 66, No. 4, pp. 275–276, April, 1989.     

Need for and requirements for a neutron irradiation facility for fusion materials testing

The construction and operation of an intense 14-MeV neutron source is essential for the development and eventual qualification of structural materials for a fusion reactor demonstration plant (DEMO). Because of the time required for materials development and the scale-up of materials to commercial production, a decision to build a neutron source should precede engineering design activities for a DEMO by at least 20 years. The characteristic features of 14-MeV neutron damage are summarized including effects related to cascade structure, transmutation production, and dose rate. The importance of a 14-MeV neutron source for addressing fundamental radiation damage issues, alloy development activities, and the development of an engineering database is discussed. For these considerations, the basic requirements and machine parameters are derived.     

含MOX燃料堆芯与传统堆芯的辐射特性对比研究

U-Pu混合氧化物(Mixed oxide,MOX)燃料应用前景广阔。以国内M310型堆芯为对象,对使用30%MOX燃料的部分低泄漏堆芯燃料管理方案进行分析,比较了含MOX燃料堆芯和传统的全UO2燃料堆芯在平衡循环下压力容器快中子注量、原子位移次数(Displacement per atom,DPA)和辐照监督管超前因子的特性差异。结果表明,与国内主流的高泄漏全UO2燃料堆芯平衡循环相比,平衡循环压力容器内表面快中子注量率和DPA率小20%左右,343°处的辐照监督管快中子注量率小8%,超前因子大15%;与国内占少数比例的低泄漏全UO2燃料堆芯平衡循环相比,平衡循环压力容器内表面快中子注量率和DPA率大40%左右。进一步分析发现,虽然同等功率下MOX燃料比UO2燃料释放的中子多7%,但与国内主流的高泄漏全UO2燃料堆芯相比,部分低泄漏MOX燃料堆芯的燃料管理方式使堆芯外围组件功率降低,使得压力容器受到的快中子辐照损伤降低。     

The role of dispersed barriers in the pulsed irradiation creep of magnetic fusion reactor materials

The effects of irradiation pulsing on the climb-glide creep, and the role of the glide barrier height are investigated. Only tokamak-type pulsing is considered. We develop a new formulation for the creep strain increment per pulse for large barriers, for which more than one pulse is needed for glide to occur. This formulation is applied to typical tokamak-type conditions, including the UWMAK-I and INTOR designs. It is concluded that no significant enhancement over steady irradiation occurs for τ_v ? burn-time. However, in long burn-time Tokamaks with τ_v?on-time and off-time, it is found that the pulsed creep enhancement can be significant. For example, for a duty factor of 0.9 the enhancement is about 3 for small barriers using a dose-equivalent average damage rate when comparing pulsed and steady irradiation. The maximum enhancements are diminished to about 2 when equal instantaneous damage rates are used.     

Evaluation of nuclear heating of small samples in a research reactor core

A simple analytical model is presented which indicates that the ratio of heating power densities of two different materials, irradiated under the same conditions inside a reactor core, can be estimated from material properties only. The developed approximate method allows simplifying the measurement technique of a number of samples from different materials by performing measurement of only one sample. The latter is a primary result of the proportionality between the heating and the thermal neutron flux for samples irradiated under the same conditions. This is confirmed by measurements in the Greek Research Reactor (GRR1) showing that the temperature of materials irradiated at various positions of a vertical core channel due to all heating mechanisms follows the spatial variation of thermal neutron flux. Validation by the numerical 3D gamma heating code GHRRC developed in NCSR Demokritos shows that the conclusions of the simple analytical method apply also for the total heating by all gammas in the core. A model is also presented for the estimation of heating by elastic neutron scattering. Furthermore, a methodology is suggested for the estimation of the temperature and the heat power deposited on materials irradiated during normal reactor operation, based on in-pile temperature measurements performed at low reactor power levels.