A unit for inspection of materials using differential gamma-ray scattering technique

The main objectives of this research were to develop a prototype unit using the differential gamma-ray scattering technique (DGST) and to demonstrate its possible use in nondestructive inspection of materials. The unit consisted of a 5 mCi (185 MBq) ¹³⁷Cs gamma-ray source positioned perpendicularly to a 5 cm × 5 cm BGO detector. The gamma-ray beam was collimated by a 5 cm thick lead collimator with 1 cm ∅ opening while the detector was only side shielded allowing scattered gamma-rays to reach the detector from different angles. The unit was then tested with 20 cm × 20 cm × 20 cm concrete mortar containing four rebars at its corners. It was found that the integral of the differential spectrum changed corresponding to the size and position of the rebar which was in front of the source and the detector. It was also found that the integral of the differential spectrum increased with increasing degree of corrosion of the rebar. The results indicated that a portable DGST unit could be designed to be used as a tool in nondestructive inspection but the interpretation of the differential spectrum still needs further investigation.     

Ultrasonic turbine shaft inspection using the synthetic aperture focussing technique (SAFT)

Ultrasonic imaging techniques developed for turbine shaft inspection result in a two-dimensional image of the defect distribution inside the specimen. It is shown, how the effect of reflectors outside of the image plane lead to misinterpretation of the reflector positions. The expansion of the two-dimensional imaging technique to three-dimensions will help to overcome these problems.     

Steam generator small bore piping socket weld inspection using the phased array ultrasonic technique

As nuclear power plants age, the likelihood of failures in the small bore piping used in those plants caused by exposure to mechanical vibrations during plant operations increases. While small bore piping failures rarely cause plant shutdown, the management of small piping has been a keen area of interest since their repair or maintenance requires a reactor trip. Steam generator (SG) drain pipe socket welds are small diameter piping connections (nominal pipe schedule 3–4 inches) susceptible to mechanical vibration. SG drain pipe socket weld failures have caused coolant leakage. Therefore, more reliable inspection technologies for small bore piping need to be developed to detect problems at an early stage and prevent pipe failures. This research aims to improve the reliability and accuracy of small bore piping inspections through the design, manufacture and application of a new phased array ultrasonic testing technique and inspection system for SG drain line socket welds.     

Qualification of the LF-eddy current technique for the inspection of stainless steel cladding and applications on the reactor pressure vessel

As part of the re-inspection of the reactor pressure vessel of the nuclear power plant, the low-frequency-eddy current technique was implemented during the 1995 outage. Since then, this inspection technique and the testing equipment have seen steady further development. Therefore, optimization of the entire testing system, including qualification based on the 1995 results, was conducted. The eddy current testing system was designed as a ten-channel test system with sensors having separate transmitter and receiver coils. The first qualification of the testing technique and sensors was performed using a single-channel system; a second qualification was then carried out using the new testing electronics. The sensor design allows for a simultaneous detection of surface and subsurface flaws. This assumes that testing is performed simultaneously using four frequencies. Data analysis and evaluation are performed using a digital multi-frequency regression analysis technique The detection limits determined using this technique led to the definition of the following recording limits for testing in which the required signal-to-noise ratio of 6 dB was reliably observed.

• Detection of surface connected longitudinal and transverse flaws:

• notch, 3 mm deep and 10 mm long, for weave bead cladding;

• notch, 2 mm deep and 20 mm long, for strip weld cladding.

• Detection of embedded planar longitudinal and transverse flaws:

• ligament of 7 mm for 8 mm clad thickness and 3 mm;

• ligament for 4 mm clad thickness, notch starting at the carbon steel base material with a length of 20 mm.

• Detection of embedded volumetric longitudinal and transverse flaws:

• 3 mm diameter side-drilled hole (SDH) for 8 mm clad thickness; ligament, 4 mm. For 4 mm clad thickness: diameter, 2 mm SDH; ligament, 2 mm. All SDHs are 55 mm deep.

重离子束的Bragg峰定位于麦粒的自由基探针技术

为了实现定点定位诱变，提出了一种利用ＥＳＲ波谱术对离子束Ｂｒａｇｇ峰在麦粒中的定位方法，即自由基探针技术，采用该技术有效地测定出２１ＭｅＶ／ｕ氮离子束的Ｂｒａｇｇ峰位于小麦胚部，注入深度不超过１．０±０．２ｍｍ，而４８ＭｅＶ／ｕ氮离子束贯穿全麦粒，不能实现定点定位诱变。     

Transmutation capabilities of the CERN Energy Amplifier System

Progress in particle accelerator technology makes it possible to use a proton accelerator to produce energy and to destroy nuclear waste efficiently. Energy Amplifier (EA) systems consist of a sub-critical fast neutron core driven by a proton accelerator. If well designed, they prevent any possible criticality accidents. It has been proposed to take advantage of this sub-criticality in order to use certain types of fuel with poor neutronic properties (for instance those with very small delayed neutron fractions). In this respect, they are particularly attractive for destroying, through fission, transuranic elements produced by present nuclear reactors. EA's could also transform efficiently and at minimal cost long-lived fission fragments using the concept of Adiabatic Resonance Crossing (ARC), an innovative method tested at CERN with the TARC experiment.     

应用于集装箱检查系统的互逆式客户/服务器处理技术

提出一种新型的分布式网络处理技术,即互逆式客户/服务器处理技术,结合该技术在“同方威视”集装箱检查系统中的成功应用,介绍了该技术的特点和优势,阐述了用这种新技术进行网络程序开发的策略和原则,并提出了用该技术进行集装箱检查系统软件开发的一个范例。     

Demonstration of the heat removing capabilities of the EPR core catcher

The EPR implements an additional, fourth level of defense-in-depth that aims at limiting and restricting the consequences of a postulated severe accident with core melting to the immediate vicinity of the plant. As this requires an intact confinement, it is necessary, among others, to avoid an attack of the molten core on the basemat. For that purpose, the EPR includes a large ex-vessel core catcher. It increases the surface-to-volume ratio of the melt after its release from the reactor pressure vessel (RPV) and allows the effective quenching and stabilization of the melt before it can attack the structural concrete.The bottom and sides of the core catcher are cooled by a system of horizontal water-filled channels. The water is provided either passively, by overflow from an internal reservoir, or actively by the containment heat removal system (CHRS).To quantify the heat removing capability of the horizontal part of the proposed cooling structure, a set of experiments in a full-scale, horizontal, 5 m long cooling channel have been performed. To simulate decay heat, the channel was electrically heated from the top. The experiment was integrated in the BENSON test rig, a highly flexible, separate-effect test facility operated by Framatome ANP. In accordance with the potential later modes of operation, both co-current and counter-current flow of the water/steam mixture have been investigated.The tests demonstrated the good-natured behavior of the system, even for induced heat fluxes that significantly exceed realistically expectable maximum values. Although, at high heat fluxes, a local dry-out occurred at the top of the channel, structural temperatures remained in a safe range. This excellent performance is attributed to the fact that heat can enter the water through both the horizontal and vertical surfaces of the cooling channel. As a result a high, effective critical heat flux (CHF) level is achieved. The performed tests yield a valuable contribution to the validation of the function of the EPR core catcher concept.     

High heat flux capabilities of the Magnum-PSI linear plasma device

Magnum-PSI is an advanced linear plasma device uniquely capable of producing plasma conditions similar to those expected in the divertor of ITER both steady-state and transients. The machine is designed both for fundamental studies of plasma–surface interactions under high heat and particle fluxes, and as a high-heat flux facility for the tests of plasma-facing components under realistic plasma conditions. To study the effects of transient heat loads on a plasma-facing surface, a novel pulsed plasma source system as well as a high power laser is available. In this article, we will describe the capabilities of Magnum-PSI for high-heat flux tests of plasma-facing materials.     

Proving test of earthquake-resistant pipings, equipment and active components

In Japan we initiated the project of a shaking table to prove the earthquake-resistant properties of key items in nuclear power stations. This two-axial shaking table will be able to shake a 1000 ton object on a table of 15 × 15 m by 2600 tonG in horizontal force and by 1300 tonG in vertical force. In this paper, the philosophy of such projects as well as various experiences on such proven tests done in Japan will be described.The main purposes of the proven tests are to understand:

1. (1)The behavior of nuclear structures and equipment under strong earthquakes both from the viewpoint of structural dynamics and process dynamics.

2. (2)The endurance limits of structures and equipment to destructive earthquakes both from the viewpoint of structural integrity and function.

3. (3)The behavior and availability of active components under deformations and accelerations induced by destructive earthquakes.

4. (4)The margin of safety of structures and equipment under assumed destructive earthquake conditions both for society at large and related engineers.

Although we have almost eleven centuries of historical data on earthquake damage, we still learn new facts from each new destructive earthquake. We have almost no experience of earthquake damage to nuclear structures and equipment. We have endeavoured to estimate the ‘modes of failure’ of various structures and items in nuclear power stations. Therefore, we should be afraid of lack of knowledge on that, because earthquakes are natural phenomena and have a somewhat unpredictable nature. However, we can understand the behavior of structures and equipment in their ultimate condition through the shaking test.One of our uncertainties on earthquake effects is the effect of vertical ground motions. The three-dimensional response seems to cause no particular problems; however, overturning and unstable moving of a solid structure are highly non-linear problems. The behavior of free-surface water in a containment under three-dimensional excitation is not known completely also. These things can be clarified only through shaking experiments using a two- or three-dimensional shaking table, including vertical component motions.The availability of active components, such as control rod driving mechanisms, valves and pumps can be checked only through shaking and/or forced deformation tests, because troubles in such active components may be induced by mechanical friction or contact of moving parts. The malfunction of electrical components is also complicated, for example chattering of the contact of relays. To evaluate the behavior of such active components and electrical components, the so-called mathematical model is inadequate. It is almost impossible to establish an adequate model of those components including all elemental factors related to its function.In Japan, we have many experiences of shaking tests of various size models and for various purposes, not only for nuclear power stations, but also for other areas. Their philosophy, methodologies, results and remarks will be briefly described.     

热释光方法在环境剂量监测中的应用

天然热释光量和年剂量是热释光断代工作中两个重要参数，它们代表着标本周边环境的放射性辐射水平。实验数据表明，环境辐射水平受到基底岩性的控制影响。经过长期改造，地表环境辐射水平将逐步减弱并匀化。     

RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320

During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is “Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)”. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed&Bleed procedure initiation. For the purpose of this, operator action with “Reactor vessel off-gas valve – 0.032 m” opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety.     

Development of a coupled neutronic/thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis

The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper.The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces.Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the usual 2 groups diffusion theory. Successively, with the usage of a developed pin-power reconstruction technique capable to account for the innovative fuel assembly design, sub-channel investigations of the individual fuel assemblies have been performed evaluating pin-wise clad temperatures. Obtained results will be discussed giving a detailed insight of the revolutionary HPLWR 3 pass core concept and understanding the physical reasons, which influence the local clad temperatures.The obtained results represent a new quality in core analyses, which takes into full consideration the coupling between neutronics and thermal-hydraulics as well as the spatial effects of the fuel assembly heterogeneity in determining the local pin-power and the associated maximum clad temperature.     

Water-cooling effects on the high-voltage and long-pulse capabilities of the KSTAR ICRF antenna

A new ICRF antenna originating from the prototype antenna was constructed for the KSTAR tokamak in 2002. The performance of the antenna was experimentally estimated at the RF test stand without a plasma. Recently three series of RF tests were performed at a frequency of 30 MHz; without any cooling, with a water-cooling for only the antenna, and with a water-cooling of the antenna and the transmission line connected to the antenna. In the tests, a half of the current strap was connected to a RF source via a matching circuit with the other half one connected to an open terminated coaxial line, and the other three straps were shorted at the input ports. During the RF pulse, the temperatures at several positions of the antenna cavity wall were measured by embedded thermocouples and the temperature profile of the front face of the antenna was measured by an IR camera. The line voltage, forward and reflected powers, and the RFTC pressure were also measured. The water-cooled antenna showed several enhanced performances in a comparison with the non-cooled case, and the standoff voltage was significantly increased. By utilizing a water-cooling of the antenna and the transmission line, we achieved a standoff voltage of 41.3 kVp for a pulse length of 300 s, and we could extend the pulse length up to 600 s at a maximum voltage of 35.0 kVp without encountering any problems, which considerably exceeds the design requirements.     

Recent status of the Kiev nuclear probe

The modified Van de Graaff accelerator with proton beam energy W ? 3 MeV has been installed and put into operation at the TMM laboratory in Kiev. The laboratory incorporates the nuclear probe (NP) beam line, coupled to this accelerator. A short version of an optimized probe-forming system (PFS) has been developed for the Kiev NP. The system is based on divided triplet of the magnetic quadrupole lenses (MQLs). This PFS has two working regimes for the probe operations. The results of numerical calculations of the geometrical and ion-optical parameters of the PFS are presented. It is shown that this versatile PFS is a promising design for a modern nuclear nano-probe. A new precision adjustable MQL has been designed. Three lenses, the slit systems and target chamber are manufactured and installed at the Kiev probe beam line. Also a new data acquisition system for the Kiev NP is being developed.