Absolute measurement of the neutron flux density in the A.E.R.E. reactor ‘gleep’

Axton, E.J.

doi:10.1016/0368-3230(63)90117-4

Cited by 28 publications

(17 citation statements)

References 14 publications

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“…The thermal neutron fluence has been calculated applying the Westcott formalism [12,13]. The final equation is given by:…”

Section: Fluence Determinationmentioning

confidence: 99%

Measurement of the thermal neutron capture cross section and resonance integral of Am-241

Maidana¹,

Dias²,

Koskinas³

2001

Radiochimica Acta

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Measurements of the thermal neutron cross section and resonance integral for the reaction 241 Am(n, γ ) 241g Am have been performed at the IEA-R1 research reactor operating at 2 MW. A 4πβ − γ coincidence system applying the tracer technique has been used for measuring the activities. Irradiations with and without cadmium cover were performed and the neutron transmission through this cover was calculated by the Monte Carlo method. High resolution γ -ray spectrometry measurements were undertaken in order to check the consistency of the results and to measure activities of the neutron fluence monitor foils. Covariance analysis has been applied to all uncertainties involved. The thermal neutron cross section for 2200 m/s neutrons and the resonance integral resulted (602 ± 9) b and (1665 ± 91) b, respectively. These results are discussed and compared with previous measurements.

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“…The thermal neutron fluence has been calculated applying the Westcott formalism [12,13]. The final equation is given by:…”

Section: Fluence Determinationmentioning

confidence: 99%

Measurement of the thermal neutron capture cross section and resonance integral of Am-241

Maidana¹,

Dias²,

Koskinas³

2001

Radiochimica Acta

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“…The efficiency of detection a of the 198Au -j-rays in the sodium iodide well-type counters was measured by counting a known activity in the gold wires, induced by exposing the wires to a calibrated flux of neutrons in the GLEEP reactor (Axton 1963) . Efficiences of 35 and 45 per cent were obtained for the counters used .…”

Section: 3 Exposure Of Rats To Slow Neutrons and Neutron Measurementmentioning

confidence: 99%

“…The flux was evaluated from the effective activation cross section & for the gold wires, the saturation y-ray count per g of gold wire R, determined from the y-ray counts, and detection efficiency e of the y-ray scintillation counter R N&a where N=number of atoms of 197 Au per g of gold . The effective cross section 6 for the gold wires was 96 barns, determined by the method of Westcott (1960) and from cadmium ratio measurements made in the GLEEP (Axton 1963) and HERALD reactors .…”

Section: 3 Exposure Of Rats To Slow Neutrons and Neutron Measurementmentioning

confidence: 99%

The Carcinogenic Effect of Localized Fission Fragment Irradiation of Rat Lung

Batchelor

Buckley

Gore

et al. 1980

International Journal of Radiation Biology and Related Studies

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In a preliminary investigation of 'hot particle' carcinogenesis uranium oxide particles were introduced into the lungs of rats either by intubation of a liquid suspension of the particles or by inhalation of an aerosol. Subsequently the animals were briefly exposed to slow neutrons in a nuclear reactor, resulting in localized irradiation of the lung by fission fragments emitted from 235U atoms in the oxide particles. The uranium used in the intubation experiments was either enriched or depleted in 235U. Squamous cell carcinomas developed at the site of deposition of the enriched uranium oxide in many cases but no lung tumours occurred in the rats with the depleted uranium oxide, in which the lung tissue was exposed to very few fission fragments. Only enriched uranium oxide was used in the inhalation experiments. Pulmonary squamous cell carcinomas occurred after the fission fragment irradiation but were fewer than in the intubation experiments. Adenocarcinomas of the lung were seen in rats exposed to uranium oxide without subsequent irradiation by neutrons in the reactor and in rats irradiated with neutrons but not previously exposed to uranium oxide. It is concluded that (i) fission fragments were possibly implicated in the genesis of the squamous cell carcinomas, which only developed in those animals exposed to enriched uranium oxide and neutrons and (ii) the adenocarcinomas in the rats inhaling enriched uranium oxide only were likely to have been caused by protracted irradiation of the lung with alpha-rays emitted from the enriched uranium.

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“…In referring to resonance neutrons, most papers [3,4] only indicate the ratio of the resonance neutrons to those of the thermal type. In actual practice it is often essential to study the characteristics of resonance neutrons in media in which their spectra differ from the simple 1/E form.…”

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confidence: 99%

Thermal and resonance neutron spectra in a graphite cube

Odintsov¹,

Koshelev²,

Malinkin³

1975

At Energy

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In view of the ever-increasing demands for greater accuracy in the measurement of neutron spectra in nuclear reactors, a great deal of attention has lately been paid to the metrological problems involved in this process [1].The creation of standard sources of resonance neutrons is an extremely important problem at the present time, yet the development of such sources is still in its infancy. In referring to resonance neutrons, most papers [3,4] only indicate the ratio of the resonance neutrons to those of the thermal type. In actual practice it is often essential to study the characteristics of resonance neutrons in media in which their spectra differ from the simple 1/E form. Clearly the study of neutron characteristics in these media would be greatly simplified if we had a standard source with a set of known spectra not obeying the 1/E law. The spectral characteristics of resonance neutrons in graphite prisms were studied in [6,7]; it was shown that, depending on the distance from the primary neutron source, the resonance neutron spectrum could be expressed in the form E-(1 + fl) where Ifll << 1.In order to create a calibrated source of thermal and resonance neutrons in the present investigation, we studied the characteristics of thermal and resonance neutrons in a graphite cube with a side of 1 m, which showed that there were a number of regions in the graphite cube in which the resonance neutron spectra differed considerably from the 1/E form and could better be expressed in the form 1/E ~(here 0.8 < ~ < 1)o The neutron characteristics studied in the graphite tube subsequently justified us in using the latter as a calibrated source of thermal and resonance neutrons with different spectral characteristics. Construction of the CubeThe cube was made up of reactor graphite blocks with a density of 1.6 g/cm 3 and placed on a scaffold so that the center of the cube lay at a height of 2 m from the floor, on a level with the center of a 28-cmdiameter sphere of 90~-enriched metallic uranium. The front face of the cube lay at 3 m from the center of the uranium assembly. The cube was placed in a room 6.5 • 7.5 • 4.5 m in size with concrete walls.Along the central axis of the cube in the direction of the assembly, an open channel 10 • 10 cm in section was provided; this was filled with graphite blocks containing cells for indicating instruments. The cells were placed along one face of the blocks in steps of 3 cm, and took the form of cylinders 20 mm in diamete r and 5 mm deep.No protective screens (cadmium or boron} were placed on the faces of the cube. Characteristics of the DetectorsIn order to reconstruct the spectra we used the reactions indicated in Table 1. The range of measurable neutron energies for the system chosen was 1.46 eV to 8.7 keV. All the materials of the indicators had a natural isotopic composition. The indium, tungsten, and copper indicators were metal foil disks 5-15 cm in diameter. The indicators for the reactions 139La(n, y) and 164Dy(n, 7) were prepared from films made of a mixture of powdered la...

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Absolute measurement of the neutron flux density in the A.E.R.E. reactor ‘gleep’

Cited by 28 publications

References 14 publications

Measurement of the thermal neutron capture cross section and resonance integral of Am-241

Measurement of the thermal neutron capture cross section and resonance integral of Am-241

The Carcinogenic Effect of Localized Fission Fragment Irradiation of Rat Lung

Thermal and resonance neutron spectra in a graphite cube

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