快变化大幅度方波信号源的研制

在快脉冲测试中,为确定测试系统的幅频及衰减特性,需要研制上升时间达皮秒级的大幅度方波信号源.介绍了其工作原理及结构组成,重点分析了在调试电路中遇到的技术难点及解决途径.信号源输出脉冲上升时间700ps,幅度在0～500V范围内可调,脉宽4ns～1μs范围内可调.     

Random thermal fatigue in fast breeder reactor: a narrow-band spectrum

Thermal fatigue crack growth in a fast breeder reactor is theoretically investigated with the aid of probabilistic fracture mechanics (PFM) under the conditions that (i) the temperature variation is a narrow-band stationary process and (ii) the crack grows owing only to the peak stress variation. First, a statistical property of residual life of the component with single crack is derived in an analytical form with the aid of an extended Markov approximation method, which is an efficient mathematical technique in PFM. Next, discussion is carried out on the generalization of the primitive model to the case with plural cracks, where a stress relaxation factor is introduced to express a stress intensity factor of each crack. Finally, a numerical example is shown to examine the quantitative behavior of the component's residual life, and sensitivity analysis is performed with respect to some model parameters.     

SNR 300 fast breeder reactor: steel containment - design, erection, testing

The Kalkar Nuclear Power Plant which is equipped with an 300 MW fast breeder reactor is being built by a Consortium mainly comprising German, Belgian and Dutch companies.The components of the fast breeder reactor are enclosed in a concrete containment which is designed to withstand severe external and internal loading.The concrete enclosure is surrounded by a steel containment which is designed to prevent the release of radioactivity following a postulated accident involving the nuclear components inside the concrete containment.The paper describes the solutions adopted for the different parts of the steel containment, the calculations verifying the suitability of the designs, the erection and the steel containment pressure and leak tests. The tests were performed with successful results in 1984.     

一种快前置放大电路的研制

简要介绍了研制的快前置放大电路,该电路是一个完全的直流放大器,具有较高的电压增益(Ar≥900),较快的上升沿(tr≤4ns),低噪声(等效输入噪声电压VN≤14μV),电路中引入贝塞尔滤波电路和反相求和电路,有效抑制了直流漂移,使电路具有极好的直流稳定性和增益稳定性,该电路主要适用于快探测器,如硅雪崩光电二极管探测器,也可用于加速器束流诊断系统,放大来自束流位置探针、相位探针的微弱信号.     

Pulsed cycle operation of a fast reactor

For many physical experiments carried out with nuclear reactors, the decisive factor is not the average output delivered over a long period of time, but rather the peak output delivered over a short period of time as required by the experiment. In such cases it appears to be most advantageous to have pulsed cycle operation of the nuclear reactor, i.e., a cycle during which the reactor output increases manyfold during a short interval of time. The advantage of work in a pulsed cycle is most apparent for a fast reactor since the lifetime of neutrons is a minimum in such a reactor in comparison with other types. This paper presents expressions for the duration of output pulses in a fast reactor, and expressions describing the time dependence of the output pulses. There are also described the possible characteristics of a pulsed fast reactor which could be used for carrying out investigations in nuclear physics.This paper was written using the results of work completed in 1956. Certain data used herein were also obtained by T. N. Zubarev.In conclusion the authors wish to express their appreciation to D. I. Blokhintsev, A. I. Leipunskoi, and O. D. Kazachkovskoi for constant attention and advice, and also to F. I. Ukraintsev and N. V. Krasnoyarov for valuable discussions.     

Sodium as a coolant for fast reactors

The results of a comparative analysis and choice of sodium as the coolant for fast reactors are presented. The facilities developed for removing impurities present in the sodium coolant and monitoring their content are described. The modeling of the mass transfer of impurities in coolants and the development of new liquid-metal coolants are examined. The results of an analysis of the anomalous situations in fast reactors, and methods for removing coolant residues from equipment and salvaging wastes are presented. It is shown that the technical solutions adopted provide reliable protection from accidents. New problems of sodium technology are formulated in application to the development of a new generation of fast reactors.     

一种低噪声快电荷灵敏前置放大器的研制

简要介绍了新型低噪声快电荷灵敏前置放大器。这种电荷灵敏前置放大器采用新的设计方法案，该前放主要采用低噪声场效应晶体管和集成运算放大器构成，其等效输入噪声≤2．2keV。该前放具有电路结构简单、体积小、输出信号上升时间快、噪声低、稳定性好等特点。     

Sodium fast reactor evaluation: Core materials

In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.     

Design of a fuel element for a lead-cooled fast reactor

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg⁻¹ of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg⁻¹ of HM.     

Flexible conversion ratio fast reactors: Overview

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. The performance achievable by the unity conversion ratio cores of these reactors was compared to an existing supercritical carbon dioxide-cooled (S-CO₂) fast reactor design and an uprated version of an existing sodium-cooled fast reactor. All concepts have cores rated at 2400 MWt. The cores of the liquid-cooled reactors are placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchangers (IHXs) coupling a primary coolant to a compact and efficient supercritical CO₂ Brayton cycle power conversion system. The S-CO₂ reactor is directly coupled to the S-CO₂ Brayton cycle power conversion system. Decay heat is removed passively using an enhanced reactor vessel auxiliary cooling system (RVACS) and a passive secondary auxiliary cooling system (PSACS). The selection of the water-cooled versus air-cooled heat sink for the PSACS as well as the analysis of the probability that the PSACS may fail to complete its mission was performed using risk-informed methodology. In addition to these features, all reactors were designed to be self-controllable. Further, the liquid-cooled reactors utilized common passive decay heat removal systems whereas the S-CO₂ uses reliable battery powered blowers for post-LOCA decay heat removal to provide flow in well defined regimes and to accommodate inadvertent bypass flows. The multiple design limits and challenges which constrained the execution of the four fast reactor concepts are elaborated. These include principally neutronics and materials challenges. The neutronic challenges are the large positive coolant reactivity feedback, small fuel temperature coefficient, small effective delayed neutron fraction, large reactivity swing and the transition between different conversion ratio cores. The burnup, temperature and fluence constraints on fuels, cladding and vessel materials are elaborated for three categories of material - materials currently available, available on a relatively short time scale and available only with significant development effort. The selected fuels are the metallic U-TRU-Zr (10% Zr) for unity conversion ratio and TRU-Zr (75% Zr) for zero conversion ratio. The principal selected cladding and vessel materials are HT-9 and A533 or A508, respectively, for current availability, T-91 and 9Cr-1Mo steel for relatively short-term availability and oxide dispersion strengthened ferritic steel (ODS) available only with significant development.     

In vessel neutron instrumentation for sodium-cooled fast reactors: Type,lifetime and location

The in vessel instrumentation of sodium-cooled fast reactors must deliver measurements that are reliable and easy to interpret over several reactor cycles in order to fulfill the safety requirements. This paper compares, with respect to this requirement, three types of detectors that are widely used in neutron measurements: fission chambers, boron-lined proportional counters, self-powered neutron detectors. We use neutron spectra that are computed for preliminary design of sodium-cooled fast reactor in different representative locations: in diluting tubes within nuclear fuel assemblies, or in the lateral neutron protections. With an evolution code, we compute the expected signal for each type of detector, to assess whether its level is sufficient, and also its evolution over three operating cycles, to examine whether it is compatible with long term measurements. The conclusion is that fission chambers are the only type able to deliver an interpretable signal for a wide dynamic of reactor power and for three or more operating cycles. The two other types are shown to be inadequate.