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1.
A walkdown of Zion Unit 2 was performed in May 1987 to verify the seismic adequacy of mechanical and electrical equipment required for hot safe shutdown. This was the first pilot walkdown performed under the SQUG guidelines. Draft Generic Implementation Procedures developed by SQUG were used by the three Seismic Review Teams to walkdown 159 pieces of equipment. Based on the walkdown, 118 pieces of equipment were found to be adequate and the remaining 41 pieces were classified as outliers. Seven of these 41 outliers were further verified by subsequent walkdown and 17 were verified by additional calculations. Anchorage repairs/improvements and minor field adjustments due to seismic interaction concerns resolved the remaining 17 cases. A trend analysis of the outliers was also performed to see if other safety-related equipment, outside the scope of SQUG, should be addressed for any shortcomings. Two trends were identified, one for the anchorage of shop fabricated instrument racks, and another for bracing of HVAC duct system supplying intake air to the diesel generators. Appropriate repairs were issued for all safety-related items affected by these trends.  相似文献   

2.
In this paper we discuss the proposals for processes which have already been realised in form of bench scale units or which have been planned, as well as those which have a high degree of development potential. A part of these cycles have in common the splitting of sulfuric acids which causes corrosion problems unsolved up to now. The essential part of the metal/metal hydride-processes is a hydrogen permeable membrane which separates the hydrogen acceptor from the water containing electrolyte melt. Actually we are intending to build up a lab cycle using a TiNi-basis membrane. The metal membranes offer a number of further interesting applications, such as (1) hydrogen production from gas mixtures at high temperatures, and (2) tritium separation from the helium of the HTR primary cooling circuit. A further promising process is the hydrocarbon hybrid cycle, in which the reduction of methanol to methane and oxygen is the key reaction. Till now we can detect a methane yield of up to 50%. An interesting combined procedure for the production of hydrogen and electricity is proposed, where sulphuric acid is decomposed by means of coal. The detailed mass and energy balance shows an efficiency of up to 57%. Thermodynamic analysis for the watersplitting cycles indicates efficiencies of up to 50%. Further research and development work is necessary in order to solve material problems and to demonstrate the suitability and availablity of the techniques using larger scale laboratory and prototype units.  相似文献   

3.
A-203 steel has been irradiated by fast neutrons to fluences of 1 × 1022 and 1.25 × 1022 n/m2. The isochronal annealing studies have been carried out from 333 to 913 K. The results show a recovery peak at ˜ 385 K, which has been explained to arise due to migration of carbon and carbon-vacancy complexes. After this stage the resistivity instead of decreasing starts increasing. This has been explained on the basis of increase of short range ordering on annealing. The influence of nickel and carbon on the recovery behaviour has also been discussed.  相似文献   

4.
The current version of the computer program SAP for the static and dynamic analysis of linear structural systems is described. The analysis capabilities of the program, the finite element library, the numerical techniques used, the logical construction of the program and storage allocations are discussed. The main advantages of the program as a general purpose code become apparent. Results of analyses as comparisons with other existing solutions are given, and running times which demonstrate the efficiency of the program are included.  相似文献   

5.
The current version of the computer program NONSAP for linear and nonlinear, static and dynamic finite element analysis is presented. The solution capabilities, the numerical techniques used, the finite element library, the logical construction of the program and storage allocations are discussed. The solutions of some sample problems considered during the development of the program are presented.  相似文献   

6.
The first phase of the SCARABEE programme has already been done with fresh fuel on single and seven pin bundles. We are describing the SCARABEE facility which includes the reactor, the sodium loop, the test section, the equipment for post mortem examination, the instrumentation and the recording systems.The methods used to determine the experimental parameters such as the power generated in the pins or the heat losses are presented.The different types and number of experiments are also described.  相似文献   

7.
A comprehensive program for severe accident mitigation was completed in Sweden by the end of 1988. As described in this paper, this program included plant modifications such as the introduction of filtered containment venting, and an accident management system comprising emergency operating strategies and procedures, training and emergency drills. The accident management system at Vattenfall has been further developed since 1988 and some results and experience from this development are reported in this paper. The main aspects covered concern the emergency organization and the supporting tools developed for use by the emergency response teams, the radiological implications such as accessibility to various locations and the long-term aspects of accident management.  相似文献   

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FIPMIGR is a computer program for studying migration of fission products in a fuel pin. Migration in a temperature gradient and in a concentration gradient is considered. The geometry is cylindrical with migration only in the radial direction.As an example the diffusion of Ba is calculated and compared with experimental results. The migration of Ba is well described using the diffusion constant for Ba in BaO and a heat of transport of −100 kJ mol−1. The great sensitivity of the theoretical prediction to temperature is clearly demonstrated. Both theory and experiments show that there is a temperature or power above which migration becomes clearly visible. The critical temperature is about 1700 K and the power level in the S176 experiments [2] was then about 40 kW m−1.  相似文献   

12.
This paper discusses the probability-based load combinations for the program dealing with the design of Category I structures, currently being worked on at Brookhaven National Laboratory (BNL) for the Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission (NRC). The objective of this program is to develop a probabilistic approach for the safety evaluations of reactor containments and other seismic Category I structures subjected to multiple static and dynamic loadings. Furthermore, on the basis of the developed probabilistic approach, a load combination methodology for the design of seismic Category I structures will also be established.The major tasks of this program are: (1) establish probabilistic representations for various loads and structural resistance, (2) select appropriate structural analysis methods and identify limit states of structures, (3) develop a reliability analysis method applicable to nuclear structures, (4) apply the developed methodology to existing Category I structures in order to evaluate the reliability levels implied in the current design criteria, and (5) recommend load combination design criteria for Category I structures. When the program is completed, it will be possible to (1) provide a method that can evaluate the safety margins of existing containment and other Category I structures and (2) recommend probability-based load combinations and load factors for the design of Category I structures.At the present time, a reliability analysis method for seismic Category I concrete structures has been completed. By utilizing this method, it is possible to evaluate the safety of structures under various static and dynamic loads. In this paper, results of a reliability analysis of a realistic reinforced concrete containment structure under dead load, accidental pressure, and earthquake ground acceleration are presented to demonstrate the feasibility of the methodology.  相似文献   

13.
The paper provides an outline on ongoing activity in Italy for the preparation and adoption of codes and guides for design, construction, and safety of nuclear plant components.  相似文献   

14.
In 1978, Commissariat à l'Energie Atomique, Electricité de France, and Novatome decided to undertake a common effort to gather a complete collection of rules to apply for design of LMFBR components. The first issue of this work is now being published by AFCEN as the “RCCM” code. The preparation of the design rules used largely the experience gained in Superphenix components analysis, and the results of the large R&D program performed as a support for the design of this plant or at longer term perspective, coordinated by a scientific advisary council of AFCEN (Association Française pour les règles de Conception et de Construction des matériels des Chaudières Electronucléaires).  相似文献   

15.
An extensive library of computer codes useful for radiation transport or shielding calculations is available from the Radiation Shielding Information Center at Oak Ridge National Laboratory. In addition to the point kernel, Monte Carlo, and discrete ordinates codes used for neutron and gamma-ray transport calculations, the collection includes cross-section libraries and codes for processing cross sections, calculating fission product inventories, proton penetration of spacecraft, electron-photon transport, and analyzing neutron activation detector data to determine spectra. A list of the most current codes is given and essential information for each is included.  相似文献   

16.
In 1972 the light water reactor safety activities conducted at the Karlsruhe Nuclear Research Center (KfK) were combined under the Nuclear Safety Project (PNS). Its primary objective was to assess in quantifiable terms the safety reserves which are provided in nuclear power plant design in a conservative approach. While in the initial phase R&D work conducted under the project was largely characterized by investigations of the design basis accidents, mainly the loss-of-coolant accident, emphasis in the past decade has been shifted more and more towards severe core and core meltdown accident analysis. The activities comprise both theoretical studies and experimental investigations, often performed in adequate, large-scale facilities. All activities have been an essential part of the reactor safety research program of the Federal Ministry for Research and Technology (BMFT) and have been coordinated with a number of other programs conducted in Germany and abroad. This paper gives a broad overview of PNS contributions to LWR safety research in the past 15 years and summarizes the results, comparing them with the general goals defined. In conclusion, the attempt is made to give an outlook on remaining activities in LWR safety research being carried out by KfK.  相似文献   

17.
The SEURBNUK-2 code is now being developed jointly by AEE Winfrith and JRC Ispra for use in Fast Reactor Containment Studies. To meet the needs of such studies and the needs of the COVA program, a number of improvements and extensions of the code have been made. A selection of these changes and illustrations of their use are given in this paper.The structural capability of SEURBNUK-2 was originally limited to the treatment of thin shells and shell junctions. Although this facility proved surprisingly useful, it was realised that a more versatile and powerful means of calculating the deformation of more complicated structural geometries would be required. The finite element code EURDYN which employs convected coordinates was adapted for the purpose, so that axially symmetric elements of the quadrilateral, triangular and thin shell families could be used to model various parts of the reactor structure. The method of coupling this finite element code to the fluid motion is described and the use of this new version of the code is illustrated and the results compared with those obtained by the original code and the ALE code EURDYN 1M. This last exercise revealed small differences between the solutions which were subsequently resolved by a further investigation involving a spherical cap problem.A feature of many reactor designs which is being modelled in the later COVA experiments is the perforated plate or porous structure. For fixed perforated plates and porous structures, the additional pressure drop and inertia effects can be included in the momentum equations by addition of suitable terms and the original technique of solution is unaltered. Details of the finite difference equations are given in this paper together with the results of check calculations which were performed to ensure the correct functioning of the code.The extensive use of SEURBNUK-2, particularly in conjunction with COVA, has highlighted a number of code problems which have been successfully resolved. Many of these related to particular circumstances, and are therefore of limited interest, but a general and quite frequent problem is that of gross bubble distortion which, if untreated, leads to logic problems within the code and consequent failure. Although the basic cause of this distortion is understood, eliminating it is not straightforward. A successful palliative is to manually rezone the bubble interface since this then avoids the logical problems in SEURBNUK with little or no effect on the calculation results. A further technique is the damping of the incipient discontinuities by automatic smoothing of the particle velocities. Examples of the use of rezoning and smoothing techniques are given.It is generally recognised that the numerical processes in an Eulerian code such as SEURBNUK introduce spurious diffusion into the solution so that pressure profiles, for example, are smoother than in an equivalent Lagrangian calculation. To give guidance to the calculator on the input parameters which affect these diffusion terms, an analysis of the truncation errors involved in the derivation of the finite difference equation was made. The various diffusion like terms are listed in this paper and their relevance to the calculation is discussed. An example is given which illustrates the changing nature of the solution as the amount of diffusion is modified.  相似文献   

18.
Significant experimental results obtained at the ROSA-IV Large-Scale Test Facility (LSTF) during the first phase of the test program (1985–1988) are summarized. The LSTF is a volumetrically scaled, full-height, full-pressure simulator of a Westinghouse-type four-loop (3423 MW thermal power) pressurized water reactor (PWR). The LSTF first-phase program investigated the fundamental PWR thermal-hydraulic responses during small-break loss-of-coolant accidents (SBLOCAs) and transients. The test matrix included twenty-nine SBLOCA tests, three abnormal transient tests and ten steady-state natural circulation tests.  相似文献   

19.
Starting from the theoretical results of the extension of Dubi's Direct Statistical Approach (DSA) surface parameter model to a cell importance model, a computer code based on MCNP has been written that directly estimates the second moment and time functions. Two versions have been developed: one that estimates the exact functions and the other that estimates the point-surface approximation functions. Preliminary results obtained using the two versions are presented.  相似文献   

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