摘要:

AbstractDue to the interest in determining the dose received by persons near a criticality accident, a method for calculating the neutron spectra and neutron dose for highly enriched uranium solutions at various hydrogen to U235atomic ratios has been developed. This method uses the output from a code for criticality calculations, and determines the average leakage neutron energy, the neutron leakage spectrum, the first collision neutron dose, and the total neutron dose. The results of these calculations show that the average energy of the leakage neutrons and the dose per incident leakage neutron per cm2decrease somewhat with increased hydrogen to U235atomic ratio, but it appears that this effect is sufficiently small so that an exact knowledge of the uranium concentration would not be necessary to obtain reasonable dose estimates. The effect of neutron scattering on the neutron spectra and doses is not evaluated. No attempt has been made to describe actual dose determination methods, since these have been adequately described elsewhere.

ISSN:0029-5639    

DOI:10.13182/NSE61-A25998

出版商:Taylor&Francis    

年代:1961

数据来源: Taylor

摘要:

AbstractThe critical parameters of coupled, fissionable assemblies in close proximity to one another are studied by differential equations for the conservation of neutrons. The coupling between the assemblies is taken into account by the use of approximate boundary conditions derived from expressions for the average neutron flow between the assemblies. Critical equations are found from the one and two group diffusion theories and from theP3approximation. The results are compared with existing experimental data. The present methods are especially suited to the study of arrays of prisms.

ISSN:0029-5639    

DOI:10.13182/NSE61-A25999

出版商:Taylor&Francis    

年代:1961

数据来源: Taylor

摘要:

AbstractThe critical parameters of coupled fissionable assemblies in close proximity are studied by a variational technique based upon an integral equation for neutron conservation. Both diffusion and transport kernels can be used, and though the former is simpler, it may lead to considerable error in certain cases. An iterative method is used to solve the resulting integral equation for the case of identical spheres. The rapid convergence implies that a constant trial function for the flux is satisfactory. It is found that limiting values are rapidly approached as the separation increases and that the effect is more pronounced if the assemblies are in a vacuum than if they are embedded in matter.

ISSN:0029-5639    

DOI:10.13182/NSE61-A26000

出版商:Taylor&Francis    

年代:1961

数据来源: Taylor

摘要:

AbstractA diffusion coefficient for hydrogen in ingot uranium was determined at 800°C on 1-in. long slug blanks. A value of 14.2×10−5cm2/sec was obtained from 15 determinations. The precision of the average value at the 95% confidence level was±3.6%. Three methods of cleaning (preparatory to analysis) were employed, no bias being detected among them. An equation was used to calculate the total gas dissolved in a sample by measuring the gas evolved after a short time and using the diffusion coefficient. In a comparison of the total evolved gas (3-hr extraction) with the calculated total gas (½-hr extraction), an error of<±2% was obtained from 15 samples.

ISSN:0029-5639    

DOI:10.13182/NSE61-A26001

出版商:Taylor&Francis    

年代:1961

数据来源: Taylor

摘要:

AbstractA Modified Zirflex process was developed in the laboratory for dissolution of 1–10 % uranium-zirconium alloy fuels clad in Zircaloy-2 to produce a nitrate solution from which uranium can be recovered by conventional solvent extraction methods. A flowsheet is presented for dissolution of 7% uranium-zirconium alloy in 5.4MNH4F-0.33MNH4NO3. Enough 1 M H2O2is added continually during dissolution to yield 0.13MH2O2in the final solution, neglecting the amount reacting. Dissolution of a 70-mil thick sample is complete in 1 hr. The solvent extraction feed is prepared by adding aluminum nitrate and nitric acid to the dissolver solution to yield a stable solvent extraction feed solution of 0.0075Muranium, 0.25Mzirconium, 1Maluminum, 2Mfluoride, and 1Mnitric acid. The off-gas is approximately 98.5% NH6, 1% H2, 0.3% O2, and 0.2% N2. Conventional stainless steel such as 309SNb or Hastelloy F appear to be suitable materials of construction with corrosion rates varying from 0.1 to 3.0 mils/month.

ISSN:0029-5639    

DOI:10.13182/NSE61-A26002

出版商:Taylor&Francis    

年代:1961

数据来源: Taylor

摘要:

AbstractA process for separating americium and curium from rare earths by anion exchange based on selective chloride complexing has been developed and tested on a laboratory scale. The separation is accomplished by sorption of americium, curium, and rare earths on Dowex 1–10X resin from a solution of 8MLiNO3followed by selective elution of rare earths with 10MLiCl and americium—curium elution with 1MLiCl. Laboratory demonstration of this process has given greater than 99.5% recovery of americium tracer containing no detectable amounts of rare earths.

ISSN:0029-5639    

DOI:10.13182/NSE61-A26003

出版商:Taylor&Francis    

年代:1961

数据来源: Taylor

摘要:

AbstractLaboratory experiments have demonstrated the chemical feasibility of incorporating soluble salts of the neutron poisons boron, cadmium, samarium, and gadolinium in solutions associated with the processing of Consolidated Edison reactor fuel (stainless steel-clad 96% ThO2-4% highly enriched UO2). At room temperature at least 0.3Mboron or neutron cross section equivalent is soluble in the 6MH2SO4decladding solution or Thorex dissolvent (13MHNO3-0.04MF-0.1MAl(NO3)3. None of the poisons were volatilized to a significant extent (i.e.,104, respectively. Countercurrent batch extractions with 2.5% TBP in Amsco resulted in concentrations of boron, rare earths, and cadmium in the uranium product which were at the limits of analytical detection, i.e., 2.5,<4, and<17 ppm, respectively. Two cycles of extraction should decrease the concentration of the nuclear poisons to acceptable levels for fuel recycle.

ISSN:0029-5639    

DOI:10.13182/NSE61-A26004

出版商:Taylor&Francis    

年代:1961

数据来源: Taylor

摘要:

AbstractIn reactor kinetics problems where it is desirable to take the space dependence of reactor flux into account, the time dependent diffusion equationmust be solved. The solution of this equation on an analog computer requires a large amount of equipment. If the equation is written the form,where the volume integral of the divergence of neutron current has been replaced by the surface integral of the current, then a large reduction in the number of mesh points and consequently in the amount of equipment is possible. An example is given and compared with the flux distribution computed on a digital machine.

ISSN:0029-5639    

DOI:10.13182/NSE61-A26005

出版商:Taylor&Francis    

年代:1961

数据来源: Taylor

摘要:

AbstractIn designing a heterogeneous reactor it is not enough to be able to calculate the nominal temperature of the fuel elements; one must be able to calculate the probability that the surface temperature is either less than a given value or lies between given limits. This paper presents a general method of analyzing this problem and applies the method to the particular case of a heterogeneous, gascooled reactor. It is shown that one need not assume each statistical variable controlling the temperature to be normally distributed; the individual variables can have any distribution. For design purposes, however, one generally must assume that any value of the parameters, between fixed limits, is equally likely, and for this case it is shown that the fuel element surface temperature itself will be adequately approximated by a normal distribution even though the independent variable has a rectangular frequency function.

ISSN:0029-5639    

DOI:10.13182/NSE61-A26006

出版商:Taylor&Francis    

年代:1961

数据来源: Taylor

摘要:

AbstractTotal neutron cross section measurements have been made on metallic samples of Ti, Zr, and Hf with the MTR crystal spectrometer. The Ti cross section was measured in the energy region 0.025 to 45 ev and values of 4.19±0.04 b and 6.09±0.13 b were obtained for the bound atom scattering and 2200 meters/sec absorption cross sections. The Zr cross section was measured in the energy region 0.02 to 78 ev and a value of 6.40±0.07 b was obtained for the bound atom scattering cross section. The Hf cross section was measured from 0.04 to 3.5 ev and a value of 102±3 b for the 2200 m/sec absorption cross section is reported. Energy values of 1.095±0.005 ev and 2.378±0.016 ev were determined for the peak cross sections of the first two Hf resonances. Values ofσ0= 6030±80 b,Γ= 0.068±0.001 ev andgΓn= 0.163±0.003 mv were obtained for the 1.095 ev resonance.

ISSN:0029-5639    

DOI:10.13182/NSE61-A26007

出版商:Taylor&Francis    

年代:1961

数据来源: Taylor