# REACTOR TECHNOLOGY REPORT NO. 15--PHYSICS

## Abstract

/ s Sections. The KAPL progrann for cross-section calculations is described. The photodisintegration of beryllium and carbon was studied with the direct interaction model. Relations were developed for (n,p) reactions and inelastic scattering. Experiments and related analyses are reported for measurement of a variety of nuclear species. The Hf/sup 174/ activation cross section was obtained from a specially enriched sample of HfO/sub 2/. Natural Hf and Dy and Dy/sup 164/ were examined by chopper techniques for the energy dependence of their absorption cross sections. Pile absorption cross sections for Mo/sup 95/, Nd/sup 143/, Nd/sup 14/s, an d Sm/sup 147/ were deduced from isotropic conversion of enriched samples in a pressurized water reactor. A redetermination of the resonance integral for cobalt was conducted with a thin sample to avoid self-shielding problems. Group Constants. A special data tape of nuclear cross sections was prepared. The tape contains the best basic information currently available for a variety of strong parasitic absorbers, fuels, structural materials, etc. Definitive experiments and calculations are reported on the determination of the diffusion length of thermal neutrons in water. A very rapid and accurate scheme for digital computation of average thermal cross sections in core life studies was developed.more »

- Publication Date:

- Research Org.:
- Knolls Atomic Power Lab., Schenectady, N.Y.

- OSTI Identifier:
- 4108493

- Report Number(s):
- KAPL-2000-12

- NSA Number:
- NSA-15-008260

- DOE Contract Number:
- W-31-109-ENG-52

- Resource Type:
- Technical Report

- Resource Relation:
- Other Information: Orig. Receipt Date: 31-DEC-61

- Country of Publication:
- United States

- Language:
- English

- Subject:
- REACTOR TECHNOLOGY; ABSORPTION; ALPHA PARTICLES; C-CODES; COMPUTERS; CRITICALITY; CROSS SECTIONS; DIFFUSION; FUELS; KARE-CODE; LIFETIME; MEASURED VALUES; MONTE CARLO METHOD; NEUTRON FLUX; NITROGEN 13; NUCLEAR REACTIONS; OXYGEN 16; POISONING; PROGRAMMING; PROTONS; REACTOR CORE; REACTORS; SHIELDING; THERMAL NEUTRONS; TRANSPORT THEORY; WATER

### Citation Formats

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```*REACTOR TECHNOLOGY REPORT NO. 15--PHYSICS*. United States: N. p., 1960.
Web.

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```*REACTOR TECHNOLOGY REPORT NO. 15--PHYSICS*. United States.

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"REACTOR TECHNOLOGY REPORT NO. 15--PHYSICS". United States.
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@article{osti_4108493,
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title = {REACTOR TECHNOLOGY REPORT NO. 15--PHYSICS},

author = {},

abstractNote = {/ s Sections. The KAPL progrann for cross-section calculations is described. The photodisintegration of beryllium and carbon was studied with the direct interaction model. Relations were developed for (n,p) reactions and inelastic scattering. Experiments and related analyses are reported for measurement of a variety of nuclear species. The Hf/sup 174/ activation cross section was obtained from a specially enriched sample of HfO/sub 2/. Natural Hf and Dy and Dy/sup 164/ were examined by chopper techniques for the energy dependence of their absorption cross sections. Pile absorption cross sections for Mo/sup 95/, Nd/sup 143/, Nd/sup 14/s, an d Sm/sup 147/ were deduced from isotropic conversion of enriched samples in a pressurized water reactor. A redetermination of the resonance integral for cobalt was conducted with a thin sample to avoid self-shielding problems. Group Constants. A special data tape of nuclear cross sections was prepared. The tape contains the best basic information currently available for a variety of strong parasitic absorbers, fuels, structural materials, etc. Definitive experiments and calculations are reported on the determination of the diffusion length of thermal neutrons in water. A very rapid and accurate scheme for digital computation of average thermal cross sections in core life studies was developed. Lattice Effects. The computer program SWAKRAUM IV is described: the mathematical and physical models employed and the several options available for calculation of thermal neutron distributions in both space and energy are included. Results are reported from PPA studies of spatial variations of thermal and epithermal neutron flux in the neighborhood of moderating and absorbing inhomogeneities. A theoretical investigation of several experiments is reported in which epithermal distributions play a decisive role. A diffusion theory routine was developed for determining gamma heating in reactors. Statistical methods were applied to the calculation of self-shielding of dispersions of absorbing particles. Experimental and calculated data are given relating tc PMA measurements with discrete, highly self-shielded absorbers and less complicated poison distributions. Reactor Kinetics. For the investigation of core instabilities, it is necessary to know the transfer function relating reactivity and neutron flux in the system. Perturbation theory results were developed for this function as part of the analysis required to described complicated, time-dependent reactor behavior induced by nonuniform variation of material properties. The technique of determining reactor subcriticality with pulsedneutron measurements was automated. A physicomathematical model was developed to study probabilities associated with occurrence of certain hazardous reactor conditions. Studies are reported on the determination of neutron generation to be expected from such phenomena as spontaneous fission of fuel, ( alpha ,n) reactions, cosmic rays, and flssion-product decay. Secular Transients. Techniques for obtaining accurate cross sections were improved and generalized for the production of nuclear data to be used with arbitrary energy-group widths in depletion studies. An N/sup 13/ positron activity in the coolant of a reactor was shown to be due to the O/sup 16/ (p, alpha )N/sup 13/ reaction. Reactor TRAM program, for three-dimensional Monte Carlo calculation of low ilexibility. One- and two-dimensional diffusion calculations of criticality, neutron fluxes, fuel and poison depletion, etc., are supplied by various components of the KARE system. Brief descriptions are given of a wide variety of calculations that are possible with KARE. A technique for generalized flux synthesis to permit increasingly accurate three-dimensional diffusion calculations is embodied in the CLAG program. Statistical problems that have arisen in connection with mathematical studies, sampling techniques,},

doi = {},

journal = {},

number = ,

volume = ,

place = {United States},

year = {1960},

month = {12}

}