Detailed Magnetic Field Measurement on a Hybrid Undulator for the JAERI FEL

A planar hybrid undulator was constructed for the JAERI PEL. In order to evaluate the performance of the undulator, magnetic field of the undulator was mapped by a detailed field measurement system. The field map obtained experimentally was compared with the map calculated by computer. The undulator field was found to be sinusoidal in the longitudinal direction and almost flat In the transverse direction. The performance of the undulator was found out to be sufficient to lase the JAERI FEL, because the irregularity of the field pattern of the undulator is sufficiently small.     

A Computational Study of a Capillary Discharge Pellet Accelerator Concept for Magnetic Fusion Fueling

An ablation-dominated capillary discharge using low atomic number elements for plasma formation to flow into an ablation-free extension barrel is a concept that provides a high energy–density plasma flow sufficient to propel fuel pellets into the tokamak fusion plasma chamber. In this concept, the extension barrel is made from a non-ablating material by coating the interior wall of the barrel with nanocrystalline diamond to eliminate mixing the propelling plasma with any impurities evolving from the barrel ablation. The electrothermal plasma code ETFLOW models the plasma formation and flow in the capillary discharge and the flow into the extension barrel to accelerate frozen deuterium pellets. The code includes governing equations for both the capillary and the extension barrel, with the addition of the pellet’s terms. It also includes ideal and non-ideal plasma conductivity models. The joule heating term in the energy conservation equation is only valid in the capillary section. The pellet momentum and kinetic energy are included in the governing equations of the barrel, with the addition of the effect of viscous drag terms. The electrothermal capillary source generates the plasma via the ablation of a sleeve inside the main capillary housing. The acceleration of the pellet starts in the extension barrel when the pressure of the plasma flow from the capillary reaches the release limit. The code results show pellet exit velocities in excess of 2 km/s for source/barrel systems with low-Z liner materials in the source for 5, 20, 45, and 80 mg pellets. The study shows that an increase in the length of both the source and the extension barrel increases the pellet exit velocity with the limitation of slowdown effects for plasma expansion and cooling off inside the barrel.     

混合能源堆包层中子学初步概念设计

提出了以天然铀或压水堆乏燃料的锆合金为燃料、轻水冷却、后处理铀钚不分离、具有良好防核扩散性能的能源堆概念。利用MCNP程序与ORIGENS程序相耦合的方法设计了包层中子学初步方案,给出了初步换料方案。能源堆对聚变堆芯参数的要求不高于国际热核聚变实验堆(ITER),可以实现利用~(238)U的目的。     

Cracking and fracture of valve shafts in the secondary steam circuit of a nuclear power plant. A case investigation.

The failure of a secondary steam circuit valve was investigated. Several fatigue cracks were initiated on valve shafts in the angle of the groove supporting the puppet. Cracks propagated obliquely toward the shaft end. The final shaft fracture occurred by shearing in valve position closed, when the shaft bearing section was diminished sufficiently. The surface of the split ring transmitting the puppet weight to the shaft was worn irregularly. Fatigue cracks initiated and propagated in the split ring groove angle due to resonance stresses at the start of the operation, when the temperature of the valve was low and the shaft steel notch toughness was also low.     

A contribution to the structure analysis of the Au(110) surface

With an extended MARLOWE version the ion backscattering spectra of K⁺ from the reconstructed Au(110) surface are calculated. The computer code MARLOWE is based on the binary collision model, relevant parameters are tested within this work. The comparison of the computed spectra with previous experiments gives excellent agreement for a missing row structure with first layer contraction of 0.25 Å and a small lateral shift of the second layer atoms by ? 0.1 Å. The analysis confirms the model proposed by Moritz and Wolf.     

Study of the behavior of the graphite-sodium system for the central rotating column in a BN-600 reactor

The characteristics of sodium permeation through graphite and the accompanying swelling of the graphite are examined for the central rotating column of a BN-600 reactor. The sodium transport parameters when sodium comes into contact with graphite at 350–500°C for up to 400 h are determined experimentally. Under these conditions, the permeation parameter is (0.13–1.3)·10⁻¹¹ m²/sec, which corresponds to an effective diffusion coefficient (0.2–2)·10⁻¹¹ m²/sec. The ratio of the increment to the graphite volume and the sodium mass there is ∼0.85. __________ Translated from Atomnaya énergiya, Vol. 101, No. 6, pp. 431–437, December, 2006.     

A study on self-terminating behavior of sodium-concrete reaction (2)

Fundamental experiments were carried out to investigate the self-termination behavior of sodium–concrete reaction (SCR). In the reference experiment, the reaction time was controlled to investigate the distribution change of Na and the reaction products in the pool and around the reaction front. The measured concentrations at the reaction front were 18–24 wt.% for Na, 22–18 wt.% for Si, and 4–3.4 wt.% for Al and Ca after the self-termination. From the thermodynamics calculations, stable materials at the reaction front comprised more than 90 wt.% solid products such as Na₂SiO₃ and no Na. Furthermore, in two sensitivity experiments with additional heating or using mortar, the concrete-ablation behavior depended strongly on the reaction at the reaction front. It was concluded that SCR termination was caused by the lack of Na at the reaction front.

The distribution of Na and reaction products could be explained by a steady-state sedimentation–diffusion model. At the early stage of SCR, the reaction products were suspended as particles in the Na pool because of the high H₂-generation rate. As the concrete ablation proceeds, they start settling down due to the decreased H₂-generation rate, thereby allowing SCR termination.     

Statistical justification for the application of a dynamic model to stationary nuclear reactors

An exact formula for the spectral density of the noise power in stationary reactors is derived on the noise power in stationary reactors is derived on the basis of the statistical theory of neutron multiplication, and takes the form of the product of the square of the absolute value of the reactor transfer function and a factor weakly dependent on the frequency.Translated from Atomnaya Énergiya, Vol. 20, No. 1, pp. 21–26, January, 1966     

Analysis of a Cylindrical Hybrid Positron Camera with Bismuth Germanate (BGO) Scintillation Crystals

Theoretical design studies of a multilayer hybrid computed tomographic system which could perform both X-ray transmission computed tomography (XCT) and positron computed tomography (PCT) are presented. Use of BGO scintillation crystals makes it possible to achieve the highest spatial resolution and coincident data collection efficiency possible with the present technology. In the new hybrid system, the multilayer positioning is achieved by use of one dimensional Anger principle with BGO crystals. The ultimate resolution and counting rate attainable with the system is discussed in reference to the true to random ratio. Present study indicates that the resolution and efficiency that can be achieved in PCT mode are 5-6 mm fwhm with efficiency gain of over a hundred in comparison to the existing NaI(Tl) ring system at UCSD.     

Stabilization of polypropylene, polypropylene blends with poly (styrene-b-(ethylene-co-butylene)-b-styrene) under irradiation: A comparative investigation

The aim of this paper is to investigate the stabilization of polypropylene in the poly (styrene-b-(ethylene-co-butylene)-b-styrene) (SEBS)/polypropylene (PP) blends under irradiation with respect to PP. The PP films, SEBS/PP films were subjected to electron beam irradiation and characterized by wide angle X-ray diffraction (WAXD), differential scanning calorimetry (DSC), gel permeation chromatography (GPC), and dynamic mechanical thermal analysis (DMTA). It demonstrated that upon irradiation, the molecular weight of PP had a pronounced decrease due to the major chain scission, and the minor chain cross-linking or chain branching occurred at the higher irradiation dose. Stabilization of PP was improved in the presence of SEBS, exhibiting an enhanced irradiation resistance.     

Assessments of the kinetic and dynamic transient behavior of sub-critical systems (ADS) in comparison to critical reactor systems

The neutron kinetic and the reactor dynamic behavior of Accelerator Driven Systems (ADS) is significantly different from those of conventional power reactor systems currently in use for the production of power. It is the objective of this study to examine and to demonstrate the intrinsic differences of the kinetic and dynamic behavior of accelerator driven systems to typical plant transient initiators in comparison to the known, kinetic and dynamic behavior of critical thermal and fast reactor systems. It will be shown that in sub-critical assemblies, changes in reactivity or in the external neutron source strength lead to an asymptotic power level essentially described by the instantaneous power change (i.e. prompt jump). Shutdown of ADS operating at high levels of sub-criticality, (i.e. k_eff 0.99), without the support of reactivity control systems (such as control or safety rods), may be problematic in case the ability of cooling of the core should be impaired (i.e. loss of coolant flow). In addition, the dynamic behavior of sub-critical systems to typical plant transients such as protected or unprotected loss of flow (LOF) or heat sink (LOH) transients are not necessarily substantially different from the plant dynamic behavior of critical systems if the reactivity feedback coefficients of the ADS design are unfavorable. As expected, the state of sub-criticality and the temperature feedback coefficients, such as Doppler and coolant temperature coefficient, play dominant roles in determining the course and direction of plant transients. Should the combination of these safety coefficients be very unfavorable, not much additional margin in safety may be gained by making a critical system only sub-critical (i.e. k_eff0.95). A careful optimization procedure between the selected operating level of sub-criticality, the safety reactivity coefficients and the possible need for additional reactivity control systems seems, therefore, advisable during the early design phase of any ADS systems in order to assure a benign transient response of the particular ADS design under investigation to typical plant transient initiators.     

A study on the dynamic behaviour of a prestressed burst protection system with varying boundary conditions

A burst protection system for a nuclear reactor signifies an additional passive safeguard. In case of a hypothetical failure of the prestressed reactor pressure vessel, this system serves to avoid the accidental release of radioactivity from the containment building. This paper contains a numerical study of the burst protection construction, including a study of the sensitivity of results to mechanical and numerical parameters. It shows to what extent an accurate evaluation of the important parameters for the numerical computation is needed when discretizing the structural system.     

Experimental investigation in the TOSQAN facility of heat and mass transfers in a spray for containment application

TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulate typical accidental thermal hydraulic flow conditions in nuclear-pressurized water reactor (PWR) containment. The TOSQAN facility which is highly instrumented with non-intrusive optical diagnostics is particularly adapted to nuclear safety CFD code validation. The present work is devoted to studying the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we perform detailed characterization of the two-phase flow.     

Modification of thermal hydraulic code,THEATRe, with fuel rod behavior models and its application to reactivity-initiated accident (RIA)

During reactor operation, many complex changes occur in fuel rod which affects its thermal, mechanical and material properties. These changes also affect the reactor response to the transient and accident situations. Realistic simulation of fuel rod behavior under transients such as reactivity-initiated accident (RIA) is of great significance. In this study, thermal hydraulic analysis code THEATRe (Thermal Hydraulic Engineering Analysis Tool in Real-time) has been modified by addition of fuel rod behavior models for dynamic simulation of nuclear reactor. Transient changes in gas-gap parameters were taken into account by modeling the gas-gap behavior. Thermo-mechanical behavior of fuel rod is modeled to take into account the thermal, elastic and plastic deformation. To simulate RIA, point reactor kinetics model is also incorporated in the THEATRe code. To demonstrate the transient fuel rod behavior, AP1000 reactor is modeled and three hypothetical RIA cases are simulated. The RIA is considered at three different reactor power levels, i.e. 100, 50 and 1% of nominal power. The investigated parameters are fuel temperature, cladding stress and strain, fuel and cladding thermal conductivity and heat transfer coefficient in gas-gap. Modified code calculates the fuel rod temperatures according to updated fuel, clad and gas-gap parameters at the onset of steady-state operation and during the transient. The modified code provides lower steady-state fuel temperature as compared to the original code. Stress and strain analyses indicate that the hoop and radial strain is higher at high power locations of the fuel rod; therefore, gap closure process will initially occur in the central portion of the fuel rod and it should be given more emphasis in the safety analysis of the fuel rod and nuclear reactor during accidents and transients.     

人工核素~(137)Cs对我国陆地γ辐射剂量率的贡献

我国陆地１３７Ｃｓ沉积所造成的γ辐射剂量率,以均匀分布模式进行估算比较合理。以此得到,我国各省的１３７Ｃｓγ辐射剂量率的均值在（７—１８） ×１０－１０Ｇｙｈ－１之间,约占陆地总的γ辐射剂量率的１％－３％;考虑到人口的分布,全国的平均值约为９×１０－１０Ｇｙｈ－１,约占总γ剂量率的１．５％左右。但最高值可达 １．７ × １０－８Ｇｙｈ－１,与其它任一天然放射性核素的贡献相接近。与此相比较,如果用张弛长度为 ３ｃｍ的指数分布模式估算,１３７Ｃｓ γ辐射剂量率的贡献比用均匀分布模式约高 １倍。     

Dose equivalent and its unit of measurement (a commentary on practical application)

Translated from Atomnaya Énergiya, Vol. 67, No. 1, p. 59, July, 1989.     

A new analysis chamber with a rotating target holder for total-sample PIXE analysis of aerosol deposits collected in Berner impactors

We describe an external PIXE analysis chamber provided with means for determining the total amount of aerosol material sampled in Berner impactors. The deposits are concentrated in equally spaced small spots along a circle with a diameter of 50 mm. To pass the whole distribution of material through the probing proton beam, the thin backing film is mounted on a rotatable target holder. Driven by a motor, the holder is rotating at a speed of typically 1/min. Depending on the total duration of the analysis, the sample is rotated 3–10 times through the beam. For high sample throughput the analysis chamber is coupled to a helium-backfilled transfer box in which six holders can be stored, at the same pressure as in the analysis chamber. Loading of the transfer box, with a minimum loss of helium, is accomplished by activating load locks on either side of the box. The first experience gained with aerosol samples as well as with dedicated reference standards is discussed with particular reference to the uniformity of aerosol deposits.     

The use of monodirectional neutron sources for the investigation of shielding (B-2 apparatus in a BR-5 reactor)

The report describes the B-2 apparatus, installed in a BR-5 fast reactor, for investigating the passage of neutrons through various shielding materials. It is shown that the monodirectional neutron disc source used in this apparatus makes it possible to obtain detailed information on the spatial-energy and angular distributions of the neutrons in the shielding. The effect of the angular distribution of the radiation leaving the source on the attenuation factor of this radiation in shielding was also investigated.In conclusion I would like to express my sincere thanks to A. I. Leipunskii for valuable advice during the formulation of the scheme of investigations concerning the passage of neutrons in different media from monodirectional sources, and I. I. Bondarenko, V. V. Orlov, V. I. Kukhtevich, Yu. A. Kazanskii, B. I. Sinitsyn, E. S. Matusevich, B. P. Shemetenko, Sh. S. Nikolaishvili, V. P. Mashkovich, and A. A. Abagyan for discussing the results of this work; and, finally, D. S. Pinkhasik 'and N. N. Aristarkhov for considerable help in making the B-2 apparatus.