ISSN:0029-5450    

DOI:10.13182/NT87-A33962

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

A dynamic reliable model has been developed to predict the steam volume fraction and the dynamics of flow regime boundaries in flow boiling. The model was established from the void propagation equation and the advanced dynamic drift-flux scheme. Numerical results showed that this model compared well with existing experimental data as well as other analytical models.

ISSN:0029-5450    

DOI:10.13182/NT87-A33963

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

Reactor physics tests at initial startup and after refuelings are performed to verify the nuclear design and to assure safe operation. Analog computers and instruments are widely used for the acquisition of data, and these data are reduced by hand. These conventional procedures, however, require much time and labor. Since there has been great progress in the development of digital computers and devices, these procedures are digitalized, which successfully reduces the time and labor required for reactor physics tests.

ISSN:0029-5450    

DOI:10.13182/NT87-A33964

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

An analysis model has been proposed to evaluate reactivity due to horizontal fast breeder reactor (FBR) core deformation in seismic events by direct three-dimensional eigenvalue calculations, which are impossible for current neutronic analysis programs. The model is based on a current-centered finite difference neutron diffusion calculation method. Macroscopic neutron reaction cross sections are defined, which take into account changes in both mesh volume and material composition. Further, the expression of vertical neutron current is modified in such a way as to take into account changes in vertical mesh interface areas. By using these macroscopic neutron cross sections and the modified expression for vertical neutron current, it is possible to calculate the effective multiplication factor of a deformed FBR core within the bounds of a finite difference diffusion calculation method using the same mesh division used for the normal nondeformed core. Computation time and computer core memory required by the presented model are almost the same as in current finite difference methods. The calculated reactivities for simple one-dimensional slab, two-dimensional slab, and three-dimensional hexagonal systems agreed within 5% of those obtained by either a finite element method or a finite difference method. The agreement was particularly good (within 2%) for cases in which fuel assembly pitches decrease around the horizontal core midplane; therefore, large reactivity is inserted.

ISSN:0029-5450    

DOI:10.13182/NT87-A33965

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

A model was developed for faster than real-time liquid-metal fast breeder reactor core transient analysis for purposes of continuous on-line data validation, plant state verification, and fault identification. The basic feature of this model is the use of a nodal approximation for the coolant, cladding, and fuel temperatures that gives adequately accurate power and temperature predictions with very few axial nodes. In applications of this methodology to fast loss-of-flow and overpower transients, computation times of about one-thirtieth of the real transient time per thermal-hydraulic channel were obtained. The predicted coolant and cladding temperature distributions were practically identical to those resulting from detailed finite difference computations. The predicted fuel temperatures differed by∼1% or less from those obtained from the same finite difference computations. The analysis of the Transient Reactor Test Facility experiment TS-1C and the Experimental Breeder Reactor II experiment SHRT-17 showed very good agreement between model predictions and measurements.

ISSN:0029-5450    

DOI:10.13182/NT87-A33966

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

To investigate the neutron dose and spectra around a fast reactor from the point of view of radiation protection and shielding, neutron measurements were conducted at the reactor top of JOYO, a Japanese experimental fast reactor, and an analysis by a transport calculation was performed. The measurements were carried out under a Mark II irradiation core with and without the reactor top concrete pit cover at 98- and 48-MW(thermal) power levels, respectively. The measurements were performed at several points in and around the reactor top pit room. Neutron detectors with well-examined response functions were employed for this study–the rem (sievert) counter as a neutron dosimeter and the multimoderator neutron detector as a neutron spectrometer. The measured neutron doses distributed from 0.4 to 100 mrem/h·[100 MW(thermal)]−1{4 to 1000μSv/h·[100 MW(thermal)]−1} and the measured neutron spectra showed an∼1/E type energy distribution. The rapid spatial change of the neutron spectrum could not be observed near the reactor top. The neutron flux distributions around the reactor were calculated and compared with the measured results. The two-dimensional transport code DOT 3.5 was employed for the calculation, and the neutron group constants were prepared by using JENDL-2 cross-section libraries. The values of measurements and calculations were in relatively good agreement within a factor of 3 to 5 in spite of the 12-decade decrease in neutron flux from the reactor core center. It is shown that the effect of stored fuels in invessel storage racks has greatly affected the neutron dose rate at the reactor top. The modeling for shielding calculations of the iron rotating plug structures is discussed.

ISSN:0029-5450    

DOI:10.13182/NT87-A33967

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

The invention at Chalk River Nuclear Laboratories of a simple method of wetproofing platinum catalysts allows them to retain their activity in liquid water. High performance catalysts for the hydrogen-water isotope exchange reaction that remain active for years can now be routinely produced. The first commercial application using the ordered-bed-type wetproofed isotope exchange catalyst developed and patented by Atomic Energy of Canada Ltd. has been successfully completed. Approximately 9100 m3of deuterium gas stored at Brookhaven National Laboratory was converted to high grade heavy water. Conversion efficiency exceeded 99.8%. The product D2O concentration was 6.7 percentage points higher than the feed D2gas.

ISSN:0029-5450    

DOI:10.13182/NT87-A33968

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

A nonlinear optimization program was used to fit the model equation to the experimental equilibrium data of a thorium-nitrate/nitric-acid-1-M di-n-heptyl sulfoxide (DHPSO) 1,1,2-trichloroethane (TCE) system successfully. The extraction equilibrium curves for this system obey Langmuir type. The product of parameters A and K is the maximum extraction distribution coefficient that can be used to compare the extraction power of 1 M DHPSO-TCE with respect to Th(NO3)4at various aqueous acidities.

ISSN:0029-5450    

DOI:10.13182/NT87-A33969

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

Data from a reactor operating with a single defective fuel element were used to develop a physically based model for describing the increased release of iodine to the primary coolant following reactor shutdown. Transport of iodine from the fuel-to-sheath gap of the element to the primary coolant is described by a diffusion process. The model has been used to predict the timing of the increased release.

ISSN:0029-5450    

DOI:10.13182/NT87-A33970

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

Previous works on cement leached in water and containing radioactive wastes like cesium and strontium agree to attribute the release in the liquid phase to a diffusion mechanism. The kinetic release can be represented by an empirical relationship in which the dependence of the leached fraction Fr = C/C0versus time t assumes the form Fr = Btnand the factor n = 0.5 is typical of a diffusion process. On the other hand, the results of our studies on cement leached in static water demonstrate that the release of calcium, considered to be representative of matrix degradation, follows a time dependence of t0.25. A model is suggested for which the release of calcium depends on superposition of two processes: a diffusion through a reaction layer of calcium silicate hydrate around the cement particles during the hydration step and a diffusion of elements from the bulk of cement toward the external surface of the specimen. Based on this schematic diffusion mechanism, some suggestions are advanced to improve the physical characteristics and to increase the retention of the radioactive waste encapsulated in the cement matrix.

ISSN:0029-5450    

DOI:10.13182/NT87-A33971

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor