The ACCEPT Monte Carlo code has been used to calculate radial dose distributions around isotropic point sources of monoenergetic electrons between 0.01 and 10 MeV in an infinite water medium. The results were averaged over beta spectra to derive distributions for 147 beta emitters of which 32 are presented. More extensive tables of distributions will be presented in a report. Distributions for monoenergetic electrons agree with recent ETRAN-code calculations of Berger and Seltzer within 2%, except at very short distances where there are differences up to several percent. Results calculated by the EGS4 code differ by up to a few percent. Distributions for beta emitters are in excellent agreement with both experimental results and ETRAN and EGS4 calculations, except at very short distances.

As the radiator for a damage-track, fast neutron dosimeter, zs7Np has advantages over 232Th, the most important being its lower energy threshold. It has not been used in personal dosimeters because of possible hazards from alpha contamination and gamma dose to the wearer. This paper describes the manufacture and properties of two types of Np radiators (as an alloy with A1 and a sintered mixture with Au) for which the removable surface activity is less than about 2 dis/min/cm2 foil. Expressions are given for the neutron detection efficiency of detectors having radiators of pure or diluted fissionable material, with or without a protective coating. These show that adequate containment of the activity need not reduce the efficiency ofa given amount of Np by more than 50%. A Np-alloy radiator (0.6 mg of Np) producing 270 tracks per rad of fission-spectrum neutrons gives a dose equivalent of 34 mrem in 2000 hr at 5 cm. Circumstances in which thermal neutrons or high energy gamma radiation may interfere with fast neutron measurements are discussed.

Calculations have been made for a D2O-moderated 252Cf assembly like that being used for the calibration of neutron dosimeters at the U.S. National Bureau of Standards and being proposed by the International Standards Organization. Leakage spectra at various distances from the assembly are given along with variations in dose-equivalent rate, average neutron energy and 235U/237Np fission ratio. The spectral shape changes rapidly near the spherical assembly and the dose-equivalent rate changes more rapidly than would be expected on the basis of the inverse-square dependence. Calibration of neutron dosimeters should therefore be made at distances greater than 15 cm from the surface. At large distances from the source, the dose equivalent per unit fluence for neutrons above 1 eV is 9.3 X 10(-9) rem cm2. The effects of the structural material, recent revisions to nuclear data files and changes in the spectrum of the source neutrons on the external field were investigated. These changes produce only about a 5% change in the neutron-dose equivalent rate. The structural material introduces negligible anisotropy in the radiation field.

Leakage neutron spectra from spheres of D,O (2-50 cm radius) containing fission sources distributed uniformly throughout the volume or located at the center of the spheres were computed using the 05R Monte Carlo code. The cross-sections of several activation (lo3Rh, "'In and "S) and damage track (237Np, r32Th and 238U) detectors were averaged over these spectra to study the change in the response of the detectors per unit dose delivered by neutrons having the calculated spectral shapes. The response of Np, Rh and In dosimeters changed by less than 10% for all these spectra. From the viewpoint of energy response, such detectors would therefore be suitable for monitoring areas in which there is a high radiation field or a risk of a critical excursion involving D,O-moderated fission neutrons.

The neutron spectrum from a 238Pu-Li source has been measured using spherical proportional counters filled with different pressures of hydrogenous gas. The spectral shape over the energy range 60 keV-1.5 MeV is in good agreement with the calculated spectrum for a similar source and the neutron output obtained by integrating the spectrum is in good agreement with long-counter measurements. The neutrons from the 238Pu-Li source have an average energy of OSOMeV, an average kerma (in soft tissue) of 1.44 nrad . cm'lneutron and 18.2 nrem * cm*/neutron.

A fast neutron personal dosimeter using the lo3Rh(n, n')lWmRh reaction has two important advantages over conventional activation dosimeters. It has a lower energy threshold and yields dose readings which are effectively independent of spectral shapes over the neutron energy range from 0.75 to 10 MeV. This paper describes the characteristics of the dosimeter in detail, shows the counting equipment associated with its usage and provides the necessary data for count-to-dose conversion. The performance of the dosimeter under laboratory and simulated criticality conditions is also presented.

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Although the fission cross section of 237Np has a "threshold" at about 200 keV, it is still significant at lower energies. To facilitate corrections of 237Np dosimeter readings for this subthreshold response, recently measured cross sections between 1 and lo5 eV have been approximated by an analytical function of energy. This was used to calculate the contribution of subthreshold neutrons to the response of Np dosimeter for measurements of fission neutrons transmitted through various materials. For most of the spectra, the effect was less than 3%, but for some it was several times larger. The fission resonance integral, from 0.5 eV to 100 keV. is 280 mb.