For successful application of boron neutron capture therapy for treatment of cancer, it is required that the tumor receives a higher radiation dose than the surrounding healthy tissue. This is achieved by the use of appropriate boron compounds with selective accumulation or retention, and the subsequent irradiation with neutrons of suitable characteristics. The basics of these two requirement are given, and the implementation in clinical trials is discussed.

The renewed interest in boron neutron capture therapy (BNCT) is driven mainly by the disappointing progress in the treatment of brain tumors by other modalities over the last decades. Even though molecular biology newer drugs and strategies may promise better results in the future, BNCT is an attractive approach. Brain tumors kill by local growth and not be metastases. Boron can be delivered to the tumor while normal brain is protected by the blood brain barrier, which can be disrupted to the degree desired. Tumor selectivity can be obtained not only by improved drug barrier, which can be disrupted to the degree desired. Tumor selectivity can be obtained not only by improved drug delivery but also by restricting the capture reaction to the region of interest by targeted radiation. Both boron drug and thermal neutrons alone are to some extent innocuous to tumor and normal tissues in this binary form of therapy. The pattern of treatment failure from uncontrolled primary tumor, the blood brain barrier protection of normal surrounding tissue and the limited range of epithermal neutrons explain why brain tumors are the main focus of BNCT research.

At therapeutic doses fast neutron radiotherapy alone gives significant tumor control can result in a significant increase in tumor control probability. As a fast neutron beam penetrates tissue it is partially moderated producing a cloud of thermalized neutrons. We describe measurements of this thermalized component using the therapy beam at the University of Washington and show how it is possible to increase its magnitude without appreciable degradation of the other clinical properties of the beam. We describe model calculation of the radiobiological effects of a boron neutron capture boost as well as in vitro cell culture measurements on the V-79 cell line, in vivo experiments on the 36B10 rat glioma system, and a preliminary experiment on a human melanoma nodule test system. Plans to extend this work to clinically-relevant tumor systems are discussed.

In Japan the boron neutron capture therapy has been applied to more than 200 patients, mostly brain tumors and some melanomas. For brain tumors, Kyoto University, Kyoto Prefectural University of Medicine, Tsukuba University and National Kagawa Children's Hospital accept patients, and for melanomas, Kobe University and Mishima Institute of Dermatological Research accept patients so far. Recently the heavy water facility of Kyoto University Reactor has been upgraded for epithermal neutron as well as thermal neutron irradiations, and for the patient treatment during the continuous operation of the KUR.

A novel application of the ¹⁰B(n, (alpha) )⁷Li nuclear reaction for the treatment of rheumatoid arthritis is under development. this application, called Boron Neutron Capture Synovectomy (BNCS), is briefly described here and the differences between BNCS and Boron Neutron Capture Therapy (BNCT) are discussed in detail. These differences lead to substantially altered demands on neutron beam design for each therapy application. In this paper the considerations for neutron beam design for the treatment of arthritic joints via BNCS are discussed, and comparisons with the design requirements for BNCT are made. This is followed by a description of potential moderator/reflector assemblies that are calculated to produce intense, high- quality neutron beams based on the ⁷Li(p,n) accelerator- based reactions. Total therapy time and therapeutic ratios are given as a function of both moderator length and boron concentration. Finally, a means of carrying out multi- directional irradiations of arthritic joints is proposed.

The use of boron neutron capture therapy in clinical treatment will require the development of intense neutron sources suitable for operation in a hospital environment. Low-energy accelerator-based neutron sources have the potential for meeting the requirements for a clinical BNCT facility. This paper describes the design, installation, and initial operation of the first high current tandem accelerator for this application. The accelerator utilizes a highly power efficient switch-mode high voltage generator to deliver proton beams with energies of up to 4.1 MeV. The use of a multi-cusp high current negative ion source in combination with magnetically suppressed accelerating tubes allows the acceleration of proton beam currents up to 4 mA. The design principles and operation of the accelerator are described.

Since 1986, the Idaho National Engineering Laboratory (INEL) has been involved in the development of epithermal neutron sources for BNCT. The INEL effort was instrumental in the implementation of an epithermal neutron beam at the Brookhaven Medical Research Reactor at Brookhaven National Laboratory. Recently, the INEL's effort has been directed toward developing advanced filter designs for use with low- power research reactors such as the 250W and 1MW class TRIGA reactors which are located at various sites and universities throughout the world. This work has focused on utilizing advanced filter materials that more effectively reduce fast neutron contamination in the epithermal neutron beam and at the same time optimize neutron economy. The INEL has also been involved in developing two concepts of producing neutron sources for BNCT using charged particle accelerators. The first concept involves the use of an electron accelerator/photoneutron source. The second concept involves the use of a charged particle beam in which the particle energy is just above the threshold energy of the reaction. This paper will review the progress made by INEL in modifying the WSU TRIGA reactor and conceptual development of an electron accelerator based photoneutron source for BNCT. The near threshold particle accelerator development will be discussed in a separate paper.

The feasibility of an accelerator-driven thermal neuron source for Boron Neutron Capture Therapy (BNCT) of skin melanoma is discussed, based on the design of an analogous source for BNCT of the explanted liver. Monte Carlo simulations were performed by introducing some variations in the moderator geometry of that source. Neutrons are generated by 7 MeV deuterons impinging on a thick beryllium target. The neutron moderator consists of a nuclear graphite structure containing a heavy water tank in turn enclosing the beryllium target. The moderator materials were chosen in order to minimize the production of prompt gamma rays. THe requested neutron fluence can be achieved in three hours with a deuteron current in the range of 3-8 mA.

In this paper we have calculated the space-dependent neutron energy spectra by changing the configuration of filter and reflector materials between the reactor core and the irradiation port. The dimensions used in this calculation are 120 cm in diameter and 250cm in depth for R-Z geometry after due consideration of the neutron irradiation facility at the Musashi reactor. The filter and reflector materials used are carbon, aluminum, alumina, silicon, iron, lead, bismuth, polyethylene and lithium-fluoride. The calculations have been performed by using the Monte Carlo code and the 2D discrete-ordinate transport code adopting the S₁₂ and P₃ approximations.

Although all of the over 200 patients treated with NCT have been irradiated by reactor-produced neutrons, in the present climate of public opinion one cannot expect the routine use of nuclear reactors in hospitals or clinics to be tolerated. While there are several other methods of producing neutrons in the necessary quantities, the Paul Scherrer Institute (PSI) is very well placed to develop techniques base don spallation, in which a beam of protons hits a production target of some convenient material such as tungsten. Because spallation produces neutrons of very similar energies to reactors, the problem of moderating them to keV energies is rather similar. However, there is an additional component of high-energy neutrons, whose intensity depends on the energy of the initial beam, which has to be taken into account in any practical design. Simulations using the well-known codes LAHET, MCNP and GEANT show that a 100 (mu) A beam of 72 MeV protons, well within the capabilities of PSI, could produce enough keV neutrons for therapy.

Boron Neutron Capture Synovectomy (BNCS) is a novel application of the ¹⁰B(n, (alpha) )⁷Li reaction for the treatment of Rheumatoid Arthritis. This potential treatment modality is in its developmental stages; in this paper results of research in two aspects of BNCS are presented. First, quantification of ¹⁰B-uptake in samples of human arthritic tissue by Prompt Gamma Neutron Activation Analysis is presented. ¹⁰B concentrations from 1625 to 2726 ppm are readily achieved. Second, ideal neutron beam studies have been undertaken and indicate that neutrons from thermal energies to 1 keV are useful for BNCS. This information is of use in designing practical therapy beams should this treatment modality be realized.

A comprehensive effort is underway to design a high- performance BNCT facility at the 2 MW University of Virginia research reactor. This endeavor includes detailed core criticality and leakage calculations, coupled neutron/photon transport analyses, and dosimetry computations. Detailed geometries are modeled with MCNP for both the core and filter, as well as for phantom dosimetry studies, whereas the symmetric and deep-penetration problem of the filter/collimeter design is solved with the DORT code. Final filter configurations are evaluated with both stochastic and deterministic methods, and the results are compared and synthesized. The complementary use of these two computational methods yields a broader insight into the problem than can be achieved by using either method alone. Calculations show that certain adjustments to the core configuration increase the leakage to the filter thereby improving beam performance. Increased performance is also achieved by strategic shaping, placement, and optimization of neutral reflectors and filtering materials in the beam tube. Results of numerous optimization studies, which led to the final beam design, are presented. Ongoing work includes integration of recently developed treatment planning codes from INEL into the dosimetry analyses. New methods of coupling discrete ordinates and adjoint Monte Carlo calculations for medical applications are also under development.

The determination of absorbed dose distribution is an essential requirement for good treatment planning in Boron Neutron Capture Therapy. In this therapy, the high non- uniformity of the spatial distribution of absorbed dose and the complexity of dose components have created the necessity of a 3D dose determination with good spatial resolution and without perturbing the radiation field with the dosimeters themselves. The goal of this work is that of attaining the possibility of constructing tissue-equivalent phantoms which act by themselves as dosimeters: after exposure they can be analyzed by means of a nuclear magnetic resonance imaging systems, and the 3D distribution of absorbed dose can be determined with good spatial resolution. the discrimination between the contributions of thermal neutrons and (gamma) - rays to the absorbed dose will be possible too.

Compact neutron sources for neutron capture therapy hold the promise of permitting wide availability for this therapeutic modality for cancer treatment. Most accelerator based neutron source concepts for this purpose are centered on (p,n) reactions using bombarding energies several hundred keV to 1-2 MeV above the reaction threshold producing high neutron yield. The neutron energies in the range of hundreds of keV to 1-2 MeV require considerable moderation and/or filtration, which reduces the output epithermal neutron flux, and fast neutron contamination is always present. Operating with proton energies closer to the threshold decreases neutron yield but allows for smaller, more efficient filters and moderators, which results in less reduction of the epithermal flux by the moderator/filter assembly. Work by this collaboration is examining the balance between total neutron yield and filter/modulator efficiency in order to achieve intense epithermal beams with low fast neutron contamination. As the first stage of this project, neutron yield and spectrum measurements on ⁹Be(p,n) and ⁷Li(p,n) reactions have been made and the results will be presented along with neutronic calculations for these systems. A radio frequency quadrupole accelerator is being used in this work.

The thick target spectrum from the ⁹Be(p,n) nuclear reaction has ben experimentally determined. The data are of use in the design of accelerator-based neutron beams for Boron Neutron Capture Therapy and Boron Neutron Capture Synovectomy due to the superior physical properties of beryllium targets compared with those made of lithium. The thick target spectrum of the reaction as a function of neutron energy and angle was measured at the Ohio University time-of-flight facility. Proton energies of 4.0, 3.7, 3.4 and 3.0 MeV were used. The neutron energy spectrum was measured using 4 lithium glass detectors at 0, 20, 40, 60, 80, and 110 degrees for 4.0 and 3.7 MeV protons, and at forward angles for 3.4 and 3.0 MeV protons. The total statistical error is less than 5 percent over most of the data set. The lithium glass detectors were calibrated using the Al(d,n) spectrum which has been accurately characterized.

Boron Neutron Capture Therapy (BNCT) has been limited by the inability to direct neutrons toward the therapeutic target and away from sensitive normal tissues. The recently developed Kumakhov lens has focused a broad incident low energy neutron beam in air to a sub-mm spot. This study examines the radiation does distribution of a converging beam passing through tissue equivalent materials. A neutron beam exiting a focusing lens is directed toward a stack of thin radiochromic media sandwiched between plastic sheets. The depth dose and beam profile within the tissue equivalent materials are determined by optical scanning and image processing of the individual radiochromic media sheets, a polymer based dosimetry medium which darkens upon exposure to ionizing radiation. The alpha particle emission from boron is examined by substituting a plastic sheet with a ⁶Li enriched lithium carbonate sheet positioned at the focal plane. The information will help determine the feasibility of applying the focused neutron beam to BNCT for therapy.

We present results on surgical technique, neutron filed and irradiation facility concerning the original treatment of the liver diffused metastases. Our method plans to irradiate the isolated organ at a thermal neutron field soon after having been explanted and boron enriched and before being grafted into the same donor. In particular the crucial point of boron uptake was investigated by a rat model with a relevant number of procedure. We give for the first time statistically significant results on the selective boron absorption by tumor tissues.

A storage facility with a capacity of 420 containers is available for the interim storage of spent fuel from power reactors at the village of Gorleben in Germany. During transportation and storage of spent fuel casks radiation exposure of the personnel is dominated by neutrons. The routine control of the dose rate limits according to the transport regulations and the licence of the storage facility is performed with conventional neutron survey meters. These monitors, calibrated for fast neutrons at radionuclide neutron sources, usually overestimate the real dose rate in unknown neutron fields. In this paper, a series of measurements with several monitoring instruments near a transport cask of the CASTOR type is presented. The results are compared with reference data for the does equivalents calculated from the measured fluence spectra using a Bonner multisphere spectrometer. Besides reliable information about neutron spectra and dose rates at the container, it was found that some of the rem counters overestimate the true dose rate by a factor of 2 or more.

The cytotoxic effect of boron neutron capture therapy (BNCT) is due to the nuclear reaction between ¹⁰B and thermal neutrons. Recently, polyethylene glycol (PEG) has attracted attention due to its ability to avoid uptake by reticulo- endothelial systems, and to accumulate in the tumor cells. We prepared boronated PEG-binding bovine serum Albumin (BSA). Prompt g-ray spectrometry showed that 250.0 +/- 4.9 ppm ¹⁰B atoms were found to conjugate to PEG-BSA. ¹⁰B concentrations in AsPC-1, human pancreatic cancer cells obtained 0, 9, 24 hrs after incubation with ¹⁰B- PEG-BSA were 0, 4.97 +/- 0.49, and 13.01 +/- 1.74 ppm respectively. It is said that the ¹⁰B concentrations between 15 and 30 ppm are necessary for effective boron neutron-capture therapy. These data indicated that the ¹⁰B-PEG-BSA were getting close to the lowest acceptable limit of 15 ppm, and when using Na₂¹⁰B₂₀H₁₈ and increasing the ¹⁰B binding site of the BSA, we could increase the ¹⁰B uptake levels in the tumors for BNCT.

Neutron diffraction is becoming an increasingly important tool for the determination of triaxial residual stress distributions for materials science and engineering applications. Indeed, virtually all of the major neutron centers in the world have, or are planning, programs in this area. In this paper the basic principles of the technique will be described, along with advantages and limitations relative to other nondestructive residual stress measurement methods. Examples of the application of the technique to sub-surface residual stress determination in metals and composites will be presented, along with some near-surface results. The significant features of new neutron instruments for residual stress study will be summarized.

Biological applications of neutron scattering almost entirely rely on methods of isotropic substitution of hydrogen ¹H by its heavier isotope ²H. The exchange of water by heavy water D₂O has been most widely used for the characterization of the internal structure of complex particles, like viruses, mostly at low resolution. This method is known as contrast variation. A more detailed structural information is obtained when small components of a complex structure have a different isotopic composition. The structural studies on ribsomes are an example which will be discussed here. A further considerable enhancement of the contrast is observed when polarized neutrons are scattered by polarized proton spins.

Residual stresses induced by welding processes can affect the integrity of structural components like tubes and pipes of thermoelectric power plants. In order to reduce or cancel these stresses, welded components are often heat treated. The residual stress field in an arc-welded 2.25Cr1Mo ferritic steel pipe was measured using neutron diffraction both before and after stress relieving heat treatment. In the first stage stresses were measured using two different kinds of neutron sources: a reactor and a pulsed one. A comparison between results obtained using G5.2 diffractometer at LLB and ENGIN equipment at ISIS will be outlined and the effectiveness of heat treatment will be shown.

The size distribution of microstructural inhomogeneities in a turbine blade sample, not operated and built of IN738 Ni- based superalloy, has been investigated by Small-Angle Neutron Scattering. Data analysis has ben performed taking into account multiple scattering, that is not negligible due to the large thickness of the sample. Results show the presence of inhomogeneities with dimensions compatible with (gamma) phase particles, already observed by scanning and transmission electron microscopy.

Phase transformations in engineering materials are inevitably related to mechanical behavior and are often precursors to residual stress and distortion. Neutron scattering in general is a valuable tool for studying their effects, and pulsed neutrons are of special value, because of the inherently comprehensive crystallographic coverage they provide in each measurement. At the Manuel Lujan neutron scattering center several different research programs have addressed the relationships between phase transformation/mechanical behavior and residual strains. Three disparate examples are presented: 1) stress induced transformation in a NiTi shape memory alloy, 2) cryogenically induced transformation in a quenched 5180 steel, and 3) time resolved evolution of strain induced martensite in 304 stainless steel. In each case a brief description of the principal result will be discussed in the context of using neutrons for the measurement.

A series of 'short' radial collimators have been implemented in the 90 degree scattering geometries on the neutron powder diffractometer at Los Alamos. The capability to perform macrostain measurements has been improved by the commensurate ability to rapidly select a sampling volume appropriate to the specimen. The compact design of the collimators was dictated by the need to fit them in a cylindrical vacuum chamber as well as providing space in which to manipulate a specimen in three dimensions. Collimators of different vane lengths were fabricated to give 4 different resolutions for which 2/3 of the diffracted intensity comes from distances of 0.75, 1.25, 2.5 and 4mm along the incident beam. Qualifying scans and a demonstration on a cracked ring, containing a steep stress gradient, are included.

Neutron Depth Profiling (NDP) is a nondestructive analytical technique for measuring the concentration of certain light elements as a function of depth near the surface of a solid matrix. The concentration profile is determined by analyzing the energy spectrum of the charged particles emitted as a result of neutron capture by the elements. The measurement sensitivity is directly proportional to the neutron beam current density. A more intense neutron beam achieved by focusing improves sensitivity for specimens of small area. In addition, a narrowly focused beam adds lateral spatial resolution to the technique, which is advantageous compared with that obtained by collimating the beam size using apertures. Capillary neutron lenses have been shown to focus a neutron beam to sub-millimeter spot size. Preliminary tests have been performed in the NDP geometry using such a focusing device. A lateral resolution in the sub-millimeter range is demonstrated by a specimen of non-uniform lateral distribution composed of a row of borosilicate glass fibers.

We have been exploring the use of small-angle neutron scattering and the method of contrast variation to give a new look at very old problem--reinforcement of elastomers by carbon black in durable rubber products. Carbon black has a hierarchy of structures consisting of particles covalently bound into aggregates, which in turn associate by weak interactions into agglomerates. We found that in one carbon black, HSA, the aggregates are rod-like, containing an average of 4-6 particles. The aggregates have another graphitic shell and an inner core of lower density carbon. The core is continuous throughout the carbon black aggregate. Contrast variation of swollen HSA-polyisoprene gels shows that the HSA is completely embedded in polyisoprene and that the agglomerates are formed predominantly by end on associations of the rod-like aggregates. The surface structure of the carbon black appears smooth over length scales above about 10 angstrom. Further studies using production carbon blacks suggest that these structural characteristics are generally present in commercial rubber composites.

Quasi-elastic neutron scattering provides a direct measure of the amount of water bound during the hydration of tricalcium silicate, the principal component of Portland cement. This can be used to follow the progress of the reaction as a function of time. To examine the effect of temperature on the process measurements were made on samples held at fixed temperatures ranging from 5 to 40 degrees C. The results show at least three different stages of the hydration reaction. The early stage appears to be controlled by surface reactivity, the late stage by diffusion through the calcium silicate hydrate gel.

Small angle neutron scattering is a unique tool in the study of the precipitation phenomena. The size, shape, number density, volume fraction and chemical composition of the precipitates can be unambiguously determined and the kinetic process can be investigated by using samples that have undergone different heat treatments or by 'in situ' experiments. Systems exhibiting slow kinetics and low precipitate phase concentration, fast kinetics for which low temperature experiments are required to 'freeze' the precipitate structure and 'in situ' experiments will be discussed. The experimental results are interpreted by kinetic models which describe the precipitation process in these diverse growth conditions.

Novel modification of the neutron diffraction method - method of Bragg diffraction angle analysis, which is routinely used for strain/stress investigations in polycrystalline materials in NPI Rez is presented. It provides sufficiently high resolution permitting investigations of both the macro- and microstains of the order 10^-4. Basic feature of the modification is the exploitation of Bragg diffraction optics - focusing in real and momentum space - using curved perfect crystal slabs. Properties of this modification are documented by results of experimental measurements.

Employing the cold neutron scattering kernel for careful mapping the micro-structure of materials lends itself on thorough knowledge of differential cross section. Using the basic quantum theory, the differential cross section for hydrogen molecules in low temperature was calculated in order to extend the method with some modifications for other liquids such as CH₄.

The neutron intensity transmitted through polycrystalline material exhibits sudden well defined increases whenever the neutron wavelength ceases to fulfil the Bragg coherent scattering condition for a specific subset of lattice planes. Since in the transmitted beam all diffracted neutrons are absent, the so-called Bragg edges build up very rapidly, because diffraction of neutrons into the full solid angle contributes to their presence. High peak-intensity neutron bursts may be detected using the current-mode technique, opening up the possibility to do both transient and stroboscopic measurements. This new method is relevant to different research areas, like transient phase transformation, melting and solidification, dynamic stress influence or materials under extreme environmental conditions. Progress has been made in rapid data acquisition and analysis as well as temperature measurement by resonance doping.

Some results are presented of strain measurements carried out using a novel, modified version of the energy- dispersive neutron transmission diffraction method which has been implemented on a conventional diffractometer at a steady-state reactor. This method, which has so far been used only on time-of-flight instruments, can also be easily and successfully used at steady-state sources. The EDNTD method presented is based on the precise measurement of the Bragg diffraction edge by using focusing crystals and is complementary to the method of strain measurements by Bragg diffraction angle analysis. It allows changes (Delta) d/d of lattice spacings with sensitivity down to some 10^-5 to be investigated for reasonable, irradiated sample volumes and counting times, even at a medium-power research reactor.

Non-intrusive inspection of large objects such as trucks, sea-going shipping containers, air cargo containers and pallets is gaining attention as a vital tool in combating terrorism, drug smuggling and other violation of international and national transportation and Customs laws. Neutrons are the preferred probing radiation when material specificity is required, which is most often the case. Great strides have been made in neutron based inspection techniques. Fast and thermal neutrons, whether in steady state or in microsecond, or even nanosecond pulses are being employed to interrogate, at high speeds, for explosives, drugs, chemical agents, and nuclear and many other smuggled materials. Existing neutron techniques will be compared and their current status reported.

This paper identifies the unique characteristics of ²⁵²Cf as a neutron source and shows how they are utilized in nondestructive testing of industrial materials and monitoring of engineering processes. Conventional use of the source to produce the thermal neutrons needed in radiography and neutron activation analysis is reviewed. The ability to directly employ the fast neutrons emitted by ²⁵²Cf is demonstrated by a number of transmission, activation and scattering methods. Techniques that use this source to assay fissionable materials are also discussed.

On-line analysis by pulsed neutron irradiation is an example of an advanced technology application of nuclear techniques, concerning real problems in the cement, mineral and coal industries. The most significant of these nuclear techniques is their capability of continuous measurement without contact and without sampling, which can lead to improved control of processes and resultant large financial savings. Compared to Californium neutron sources, the use of electrical pulsed neutron generators allows to obtain a higher signal/noise ratio for a more sensitive measurement, and allows to overcome a number of safety problems concerning transportation, installation and maintenance. An experiment related to a possible new on-line raw material analyzer is described, using a pulsed neutron generator. The key factors contributing to an accurate measurement are related to a suitable generator, to a high count rate gamma ray spectroscopy electronics, and to computational tools. Calculation and results for the optimization of the neutron irradiation time diagram are reported. One of the operational characteristics of such an equipment is related to neutron flux available: it is possible to adjust it to the requested accuracy, i.e. for a high accuracy during a few hours/day and for a lower accuracy the rest of the time. This feature allows to operate the neutron tube during a longer time, and then to reduce the cost of analysis.

A method for element analysis is proposed which is based on measurements of energy changes in fast neutron scattering. Liquid scintillators (NE213 and NE230) are considered as neutron spectrometers for these energy measurements. Lineshapes of NE213 and NE230 detectors for monoenergetic 7.5 MeV neutrons have been determined and the response functions of the NE230 detector to neutrons have been determined and the response functions of the NE230 detector to neutrons scattered at 150 degrees from samples containing the elements C, N, O, Si, Al and Fe have been measured. A method for estimating C:N:O atomic ratios from scattering measurements on compounds containing these elements is described and discussed.

The Idaho National Engineering Laboratory is using the passive active neutron (PAN) radioassay system to assay 208 liter drums for plutonium in transuranic waste. PAN systems similar to INEL's are in use to perform like tasks at other US DOE sites and later versions of the PAN are in use throughout the world. The PAN system, developed by Los Alamos National Laboratory, uses both passive neutron coincidence counting and thermal neutron interrogation/differential die-away to determine the amount of plutonium in a waste drum. This paper discusses a modified statistical sampling and verification approach to determine the total uncertainty of a PAN systems measurement when applied in the assay of real waste. In this approach a statistical review is performed on real waste drum radiographs, waste generator data and PAN assay data to identify the variations of the physical, chemical and measurement characteristics pertaining to a waste category. Using data from the statistical review, the performance of the PAN assay systems is simulated using computer models of the assay system with regard to the waste form under review. The simulated responses, in terms of plutonium mass in these cases, are compared with the known input masses to assess total uncertainty. To demonstrate the method, the result of this approach will be presented for PAN measurements on two waste categories at INEL.

The Idaho National Engineering Laboratory (INEL) is developing a novel photoneutron-based nondestructive evaluation technique which uses a pulsed, high-energy electron accelerator and gamma-ray spectrometry. Highly penetrating pulses of bremsstrahlung photons are produced by each pulse of electrons. Interrogating neutrons are generated by the bremsstrahlung photons interacting within a photoneutron source material. The interactions of the neutrons within a target result in the emission of elemental characteristic gamma-rays. Spectrometry is performed by analyzing the photoneutron-induced, prompt gama-rays acquired between accelerator pulses with a unique, high- purity germanium gamma-ray detection system using a modified transistor reset preamplifier. The detection system, the experimental configuration, and the accelerator operation used to characterize the detection systems performance are described. Using a 6.5-MeV electron accelerator and a beryllium metal photoneutron source, gamma-ray spectra were successfully acquired for Al, Cu, polyethylene, NaCl, and depleted uranium targets as soon as 30 microsecond(s) after each bremsstrahlung flash.

A continuous spectrum of neutron switch energies up to 8.2 MeV is produced by a 4.2-MeV nanosecond-pulsed deuteron beam slowing down in a thick beryllium target. The spectrum form the locally shielded target is collimated to a horizontal fan-beam and delivered to a row of 16, 6-cm square plastic scintillators located 4 m from the neutron source. The scintillators are coupled to 12-stage photomultiplier tubes, constant-fraction discriminators, time-to-amplitude converters, analog-to-digital converters, and digital memories. Unattenuated neutron-source spectra and background spectra ar recorded. Luggage is stepped through the fan beam by an automated lift located 2 m from the neutron source. Transmission spectra are measured, and are transferred to a computer while the location is advanced one pixel width. As the next set of spectra is being measured, the computer calculates neutron attenuations for the previous set, deconvolutes attenuations into projected elemental number densities, and determines the explosive likelihood for each pixel. With a time-averaged deuteron beam current o 1(mu) A, a suitcase 60-cm long can be automatically imaged in 1600s. We will suggest that time can be reduced to 8s or less with straight-forward improvements. The following paper describes the explosives recognition algorithm and presents the results of teste with explosives.

The HIsmelt Process is a direct smelting process involving a coal-injected smelting reactor with the off-gas passing through a circulating fluidised bed in which fine iron ore is pre-reduced and pre-heated. Control of the pre-reduction system would benefit from an on-line measurement of the degree of pre-reduction (PRD). However, the on-line measurements need to be made on hot ore in circumstances where access to representative sample streams is very difficult. An on-line gauge based on the measurement of neutron inelastic scatter gamma-rays has been installed and calibrated at the HIsmelt plant in Kwinana, Western Australia. The gauge uses an ²⁴¹Am-Be neutron source and a BGO detector inside a water-cooled tube surrounded by hot iron ore. THe gauge monitors the iron, oxygen, carbon and silica contents of the ore based neutron inelastic scatter gamma-rays at 0.85, 6.13, 4.43 and 1.78 MeV respectively. In plant trials at HIsmelt the gauge determined iron, oxygen, carbon, silica and PRD to within 2.2 wt. percent, 0.6 wt. percent, 0.5 wt. percent, 1.0 wt. percent and 0.4 percent respectively. Potential application so this technology in other direct reduction and smelting processes will also be discussed.

A neutron transmission spectrometer has been used to determine the presence of contraband substances in sealed containers. A pulsed 'white' neutron source was created by allowing a pulsed beam of 5 MeV deuterons to impinge on a thick beryllium target. The neutron intensity was measured from about 0.8 MeV to about 4 MeV, first with the sample out of the beam and then with the sample in the beam to determine the neutron attenuation as a function of neutron energy. The neutron attenuation can be considered as the signature of the sample and can be used by a trained neutral network to identify contraband directly. The neutron attenuation can also be used along with the total cross sections to determine the number density of H, C, N and O in the neutron beam. These number densities can be used in a trained neutral network to determine the presence of contraband. Previous work used collimated neutron beams to determine the signatures of many substances individually and when imbedded in suitcases. This work is now being extended to include the complete suitcase using a a cone beam of neutrons and an array detector which covers the complete suitcase.

Energy-dependent neutron attenuations measured by time-of- flight techniques are fitted with measured cross sections to determine projected number densities of H, C, N, and O in each 3 by 3-cm² pixel of a suitcase. Attenuations due to other elements are lumped together as element X. Number densities are normalize by the total to reduce effects of variable thickness. Normalized number densities are constrained to lie in a 4D space. That space is finely binned. A fifth dimension contains six bins of N + O density, as deduced from average suitcase thickness and measured projected number densities. Each bin is labeled with the probability of encountering an explosive, as determined from distributions of elements from simulated explosive and nonexplosive situations. Probabilities for individual pixels are combined to produce an overall probability of an explosive suitcase. About 130 measurements involving 50 suitcases have been made. For about half of those measurements, an explosive was placed in the suitcase. Ten different plastic explosives were used in varying amounts and configurations. When suitcases are ordered by explosive probability, and a threshold is set to minimize errors, the false alarm rate is about 2 percent and the explosive detection rate is about 88 percent. Potential improvements are discussed.

The quantification of transuranic material (TRU) in waste packages is a common problem of countries working in the field of nuclear nondestructive inspection. The direct measurement of TRU mass inside large size closed containers is difficult due to several effects, mainly the matrix attenuation and uncertainty on the localization of the radioactive mass. The present document describes the method being developed to assay conditioned waste packages using a transportable linear accelerator which is called Mini- Linatron. The system uses a pulsed electron beam from the Mini-Linatron to produce high energy bremsstrahlung photon bursts from thin metallic converter. The transportable linear accelerator operates at 6, 9 and 11 MeV with a repetition rate between 10 to 300 Hz and a 4.5 microsecond(s) pulse duration. The maximum gamma dose rate is about 23 Gy/mn at 1 m. The photons induce fission in fissile and fertile nuclei. We counted delayed neutrons emitted after each pulse by using Sequential Photon Interrogation and Neutron Counting Signatures technique which has ben developed in this framework. Results of measurements on an experimental active gamma interrogation facility for embedded intermediate and low level wastes are presented. Finally, the advantages and performances of the photofission interrogation technique, the use of a transportable electron linear accelerator as a particle source, and the experimental and electronic details will be discussed.

Non-destructive interrogation systems designed to locate illicit substances in sealed containers involve decision making when the available objective information is incomplete. The greater the quantity of information, the more reliable is the determination of the unknown content. Therefore, it is important to be able to utilize all possible measured data pertaining to the unknown object. Among the data which can be considered are x-ray measurements, fast-neutron transmission measurements, cargo manifest data and, possibly, information of a physical, chemical or even psychological nature. The Bayesian approach provides a statistical framework for merging diverse information about any object, including a priori knowledge, subjective knowledge and objective knowledge gained from current measurements. This paper outlines the fundamental principles of Bayesian analysis and explores possible applications to the detection of illicit substances.

A proof-of-principle (POP) contraband detection system (CDS), is under construction at Northrop Grumman's Advanced Technology and Development Center. We employ gamma resonance absorption to detect nitrogen or chlorine in explosives and certain forms of illegal drugs. Using tomography, 3D images of the total density and selected element density are generated. These characteristics together may be utilized with considerable confidence in determining if contraband is present in baggage or cargo. The CDS employs a high current DC electrostatic accelerator that provides a beam of proton sat either 1.75 or 1.89 MeV. These high energy particles impinge upon a target coated with ¹³C or ³⁴S. The resultant resonant gamma rays are preferentially absorbed in either ¹⁴N or ³⁵Cl. Because of the penetrating power of the gamma rays, this approach can be utilized for inspection of fully loaded aircraft containers such as the LD3. With proper design for handling baggage and with certain CDS design features not present in the POP, throughputs approaching 2000 bags/hour are estimated. Furthermore, the technology being developed for the CDS accelerator may have other applications such as for Boron Neutron Capture Therapy or Neutron Radiography. Our current program calls for testing of the POP CDS by late 1996. This paper presents the overall design, and characteristics of the CDS POP and the status of the program.

Vehicle and Cargo Inspection System (VACIS) is comprised of a 1 Curie ¹³⁷Cs gamma-ray source in a shield and collimator which produces a fan-shaped beam designed to intercept a vertical array of gama-ray detectors contained in a tower structure. The source and detector modules straddle the vehicle or container being inspected and are mounted on self-propelled trolleys which travel in synchronization along two parallel tracks covering the length of the scanned object. The signals from the gamma-ray detector array are processed and displayed so as to produce a 2D gamma-radiographic image of the object. Testing of the system on a variety of empty and lightly-loaded vehicles and containers has demonstrated the effectiveness of VACIS in detecting hidden contraband. For example, a small sample of cocaine only 1.5 inches thick was readily detected in a container.

Fast-Neutron Transmission Spectroscopy (FNTS) is being investigated for detection of explosives in luggage or other small containers. This technique uses an accelerator to generate nanosecond-pulsed deuteron beams that strike a target, producing a white source of neutrons. Elemental distributions along projections through the interrogated object are obtained by analyzing neutron transmission data. Tomographic reconstruction is used to determine the spatial variations of individual elemental densities are combined in a detection algorithm that indicates the presence or absence of explosives. The elemental unfolding and tomographic reconstruction algorithms have been validated by application to experimental data. System studies have been performed to study the operational characteristics and limitations of a FNTS system, and to determine the system's sensitivity to several important parameters such as flight path length and the position of the interrogated object.

Radiation protection dosimetry comprises the definition of quantities to be measured, the realization of the units and the measurement of radiation in practice. Some principles of measurement for neutron protection are reviewed, the present state of the art in neutron dosimetry is described and an outlook is given on future development.

The existing cavity theories do not consider situations where the ranges of secondary heavy particles exceed the thickness of the wall of the ionization chamber. The cavity theory was modified, by introducing the wall correction coefficients, (beta) , to account for the influence of the ionization chamber wall material. The values of the coefficients (beta) were found experimentally in high energy neutron beam generated at 660 MeV phasotron in Dubna. A thin walled ionization chamber with replaceable radiators made of different materials was used for this purpose. The absorbed dose was determined in dependence on depth in a water phantom. Some necessary dosimetric parameters like the energy expended to create an ion pair and the restricted stopping power ratios were determined using correlation between these quantities and the recombination index of radiation quality. The accuracy achieved in measurements using a tissue equivalent chamber is estimated to about 4 percent at depth of 12 cm.

This paper gives a review on the current knowledge of neutron exposure to men at civil flight altitudes due to neutrons, and presents recent experimental and calculation data of the authors. The ICRP proposed in 1991 to consider air crews and other frequently flying persons as occupationally exposed people. As neutrons contribute the major part to the radiologically relevant dose at civil flight levels to men, special considerations are necessary with respect to the required basic data for the determination of the risk relevant radiation doses. The quantification of the dose, suffers from spectral data available especially in the energy range above 20 MeV. Experimental data were obtained at a low flight level on top of the mountain Zugspitze at 3000 m using a modified Bonner sphere spectrometer. The resulting spectra are compared with Monte Carlo transport calculations from top of the atmosphere down to 700 g/cm². These data and others from the literature are used to calculate the operational and risk related quantities, i.e. ambient dose equivalent, effective dose and effective dose equivalent. The question of the proper choice of the operational quantity ambient dose equivalent to estimate the risk related doses is elucidated.

A brief review is given of results obtained with the PTB Bonner sphere spectrometer, emphasis being placed on metrological and dosimetric applications of the instrument.

The neutron field around a POLLUX cask filled with nuclear waste was simulated by using a reduced model cask filed with a linear ²⁵²Cf source. MCNP transport calculations and Bonner sphere measurements were performed at different positions around the cask placed in various environments. The calculated and measured neutron spectral fluences are in remarkably good agrement.

A new design for a position-sensitive detector system for thermal neutrons is introduced. The detection principle with a thin ⁶LiF converter on the surface of a semiconductor diode is described. In experiments with thermal neutrons, a spatial resolution of 1.25 mm was obtained. The detector is insensitive to (gamma) -rays with energies up to 1.5 MeV. The design of a detector with an improvement of the detection efficiency for thermal neutrons from 2.5 percent up to 35 percent is also proposed and the present state of the process development for its fabrication is described.

'Monoenergetic' neutron fields are needed for determining the energy-dependent response of neutron dosimeters and detectors. The spectral neutron fluence is experimentally determined by time-of-flight spectroscopy with well- specified NE-213 scintillation detectors. These experimentally determined distributions are compared with calculations simulating the neutron production, attenuation and scattering processes in the target as well as energy and angle straggling of the impinging charged particles. Moreover, the diffusion of reactive nuclei into the backing is accounted for.

It is shown that fibers doped with (Ge + P) or with (GE + B) can be used with good accuracy for fluence determinations of fast or slow neutrons, respectively, when they are calibrated at sources of known flux. Calibration at easily available gamma sources, as well as fast neutron dose determination, are complicated by the facts that energetic recoil protons generated by fast neutrons in H-containing fiber coatings contribute to the dose in the fiber core, and that the highly ionizing secondary particles released by fast and slow neutrons will cause significantly lower light absorption than beta or gamma rays of the same dose.

Dosimetric characteristics of neutron and photon components of the mixed field have been determined at those four work places of nuclear power plant Dukovany where neutrons are of concern. The results were compared with data obtained at the nuclear power plant Ringhals and at the Central interim storage facility Oskarshamn, Sweden.

Within the framework of a research program of the Commission of European Communities the reference radiation facility at CERN provided recently a mixed high-energy radiation field simulating the one found at civil aviation altitudes of about 10 km. In order to calibrate radiation detectors and to investigate their response to high-energy neutrons a series of measurements using five different instruments equipped with low pressure tissue-equivalent proportional counter (TEPC) and one recombination chamber (REM-2) has been carried out. The reference radiation field at CERN was produced by reactions of a primary beam mixed of 2/3 protons and 1/3 pions with a copper target and moderated by a concrete shielding. The paper summarizes the results obtained with the five TEPC instruments and the REM-2, and discusses their characteristics.

The 'HERBE' is new coupled fast-thermal core, designed in 1991, at the 'RB' reactor in the 'Vinca' Institute. It is used for verification of designed oriented computer codes developed in the Institute, training and sample irradiation in fast neutron field. For the last purpose a vertical experimental channel (VCH) is placed in the central axis of the fast core. Neutron spectrum in the center of the VCH is calculated in 44 energy groups. Space distributions of two energy group neutron flux in the 'HERBE' are measured using gold foils and converted into the neutron absorbed dose using group flux-dose conversion factors. Gamma absorption doses in the air in the center of the VCH are measured using calibrated small ionization chamber filed with air. Determined dose rates are related to the reactor power. The first preliminary irradiations of silicon diodes in the center of the VCH of the 'HERBE' fast core are carried out in 1994 and 1995. This paper describes calculation methods and measurement techniques applied to determination of the irradiation performance and dosimetry characteristics of the 'HERBE' system.

Neutron dose calculations were performed within the framework of the 'Atkives Handhabungsexperiment AHE', which is part of the Research and Development program on the German Final Repository. The neutron dose fields of the German Pollux cask for disposal of spent fuel were calculated above ground and in a salt mine with spent fuel loadings and with a Cf²⁵² source. In an experiment performed in a former salt mine with a small shielding cask and a Cf²⁵² line source, the future exposure of the mining personnel can be estimated by measuring the neutron dose field. The neutron dose calculations ion this smaller shielding cask predicted the following experimental results and guaranteed that the neutron dose field of this smaller cask was able to simulate the realistic conditions with the POLLUX cask. They further revealed an important neutron dose rate resulting from backscattered neutrons on the salt drift walls.

The SCK CEN provides a calibration service for neutron monitors using a ²⁵²Cf source located in a facility for nuclear calibrations. Since the determined calibration factor should not depend upon the irradiation facility, corrections are to be made for neutron scattering. There are generally three techniques to correct for the effects of scattered neutrons: the semi-empirical, the polynomial fit and the shadow cone technique. In the calibration facility of the SCK CEN the source is located quite near a corner of the room, giving rise to a non-uniformity of the scattered component. This study was set up to investigate the effect of this non-uniformity on the validity of the three correction methods. A Studsvik 2202D neutron monitor was used for the experiments. The results were also compared with the results from a numerical modellisation of the neutron transport with the TRIPOLI-3D code.

High-energy spallation neutron sources are now being considered in the US and elsewhere as a replacement for neutron beams produced by reactors. High-energy and high intensity neutron beams, produced by unmoderated spallation sources, open potential new vistas of neutron radiography. We discuss the basic advantages and disadvantages of high- energy neutron radiography, and consider some experimental results obtained at the Weapons Neutron Research facility at Los Alamos.

We have developed a neutron detection system for accelerator based neutron radiography and tomography based on a combination of scintillation screen and large aperture optics combined with a cooled CCD camera. The system is capable of detecting single neutron events and can therefore be considered as a possible detector for neutron scattering as well as conventional imaging. The system has a resolution of 0.1 mm or 1242 by 1152 pixels. The limit of image size is set by the light output of the scintillator, the light collection of the optical system, the size of the CCD and the desired signal to noise ratio. The lower limit on neutron flux is determined by the dark current of the chip. Equations for these limits have been derived and can be used to predict and optimize performance. The scintillation light output per incident neutron is large enough to permit the use of lens coupled systems with their increased flexibility and ease of implementation. The system can approach a quantum limited noise level, depending on the particular geometry used. For our current system, based on the use of NE 426 scintillator,³ a 1242 by 1152 pixel EEV CCD operating at -50C, and using a 100 mm focal length, F/0.9 lens, the maximum size for the imaging screen is 0.5 m, and the lower limit for flux is 1 n/pixel/s based on this size screen and a typical dark current of 10 e/pixel/s. We are now investigating a new type of imaging technology based on large amorphous silicon sensor arrays being developed by Xerox and others. A typical device is 200 by 250 mm with a pixel size of 127 micrometers and the entire array with all electronics is in a 400 mm by 37 mm package. Major advantages of this device are the high light coupling between scintillator screen and the sensor as well as the more compact nature of such an array, since no lens systems is required and, potentially, a much lower cost. Currently, the noise performance is worse than that of CCDs, largely due to the current electronic readout but should be adequate to produce quantum limited images if electronics can be improved.

Among neutron imaging techniques, computed tomography represents the most powerful fool when an accurate 2D mapping of the internal structure of an object is desired. In order to exploit the capabilities of this technique, it is necessary to develop a detection system able to acquire the required projections with an enhanced spatial resolution and sensitivity. This could be achieved by the use of novel CCD sensors, but they are not yet available in commercial cameras. Thus, a modular system for the readout of CCD sensors was constructed allowing a fast data acquisition and a simple camera that could be easily upgraded by simply replacing the active sensor has been built. The entire system was used to acquire projections for neutron computed tomography, offering good performance spatial resolution and data transfer rate. A numerical evaluation of system performance when performing NCT is given by the analysis of sample objects.

An initial design for a high-resolution, transportable neutron radiography system concept has been developed.The system requirement is taken to be a thermal neutron flux of 10⁶ with a L/D of 100. The approach is to use an accelerator-driven neutron source, with a radiofrequency quadrupole (RFQ) as the primary accelerator component. Initial concepts for all of the major components of the system have been developed, and selected key parts, such as the RFQ and target-moderator subsystems, have been examined further. The concept features a compact trailer-mounted configuration during transport, with deployable neutron production and radiographic imaging assemblies. An overview of the system design is presented, together with brief descriptions of the concepts for each of the major subsystems. Studies completed for the RFQ and moderator are summarized, and issues identified during the course of the work are discussed.

Transportable accelerator based sources are the only means by which large structures such as aircraft or industrial components can be inspected practically using neutron radiography. Since such sources are generally considerably lower in flux than reactors, the entire system must be designed for high imaging efficiency. We describe an imaging system for accelerator based thermal neutron tomography and radiography based on the use of a radiofrequency quadrupole (RFQ) using a Be target. Fast neutrons, produced through the reaction ⁹Be₄(d,n)¹⁰B₅, are moderated with typical thermalized neutron fluence rates of 10³ to 10⁴ n/cm²/s, depending on the collimation. The camera system consists of a 1242 by 1152 pixel CCD cooled to -50 degrees C viewing a 20 by 20 cm NE-426 scintillator through an F/0.9 100 mm focal length lends. Our experiments show that neutron statistical noise and not the performance of the CCD system is the dominant limitation in the camera. We have used the system for imaging corrosion in lap joints of KC-135 aircraft and have been able to quantify corrosion within the joint.

We describe an integrating cathode-pad read-out wire chamber for imaging high energy neutrons. This technology allows construction of a detector with a wide dynamic range, which is gatable, easily read out, and provides millimeter scale resolution. Results from a prototype 48 by 48 pixel array are presented.

A computational model has been developed to investigate and assess possible applications for the technique of fast neutron radiography. This technique has a number of potential advantages including the ability to use a neutron source of high brightness, provided by an accelerator, and the use of the resonant structure of the total neutron cross-section to enhance the contrast. The development and implementation of the model is described and a hypothetical example is given of its application to the imaging of aluminium in alumina. In this case the model predicts that the signal-to-noise ratio could be increased by a factor of ten by decreasing the thickness of the lithium target from 150 micrometers to 50 micrometers for the reaction ⁷Li(p,n)⁷Be used, with 4.4 MeV protons, as a neutron source. The model is at present being tested experimentally.

Due to the special interest in applications on radiotherapy, radiobiology, analysis and material testing, in the last years the Bucharest U-120 classical variable energy Cyclotron was employed as an intense source of fast neutrons produced by 13.5 MeV deuterons on a thick beryllium target, mainly for radiobiological and archaeometrical studies. Energy spectra and yields, average energy and irradiation doses were determined using time-of-flight, multiple foils and thermoluminescent detector methods. Studies on liver chromatin structure modifications and on some archaeological Greek objects; composition using FNAA are presented.

Sarcopenia, defined as the loss of skeletal muscle with age, may lead to frailty, fractures due to falls, and reduced immunity to disease. By understanding the causes of muscle loss with age we will be able to develop ways of maintaining functional capacity and quality of life for the elderly. Elemental Partition Analysis (EPA) is a new approach to body composition assessment. A major element of the body is measured and then, by means of other measurements, is partitioned to the contributing body compartments. We developed a model for measuring total body muscle by applying the EPA method to total body phosphorus (TBP). We measure TBP by in vivo fast neutron activation analysis using the reaction ³¹P(n,(alpha) )²⁸Al. The main contributors to TBP are bone and skeletal muscle. Adipose tissue and the liver contribute less than 3 percent. We use dual-energy x-ray absorptiometry (DXA) to evaluate the contribution of bone to TBP. COrrections are applied for the small contributions of the liver and adipose tissue to TBP to derive muscle phosphorus. The technique requires high precision measurements for both TBP and DXA. The total body radiation exposure for measuring human subjects is 0.30 mSv.

A facility was constructed for the in vivo measurement of phosphorus in the bones of the human hands. The technique is based on the detection of the delayed (gamma) -rays emitted from ²⁸Al produced via the ³¹P(n,(alpha) )²⁸Al reaction. The 4-min irradiation by two ²⁴¹Am-Be sources is followed by a 4-min counting of the induced activity with two Nal(Tl) detectors. The long-term in vivo precision of the measurements was found to be 2.8 percent and the effective dose is 50 (mu) Sv. To date, well over 1000 measurements have been carried out. The amount of phosphorus in hands in healthy adults was correlated with sex, age, height, forearm bone mineral content, bone mineral density in the spine, total body fat free mass and total body cell mass. Moreover, the measurement of phosphorus in the bones of the hands of patients provided answers to a number of clinical problems.

A neutron facility suitable for radiobiological applications is described. The small chamber can house, either, solid targets or a gas target. Using this facility, absorbed doses ranging up to 7.2 Gy were delivered to Drosophila melanogaster larvae in order to study induced somatic mutation and mitiotic recombination. Some preliminary results concerning these effects, related to a mean neutron energy of 2.15 MeV provided by a d + Be source, are presented.

A prompt-(gamma) neutron activation analysis facility for in vivo determination of total body protein (TBP) in rats has been designed. TBP is determined in vivo by assessment of total body nitrogen. The facility is based on a ²⁵²Cf radionuclide neutron source within a heavy water moderator assembly and two NaI(Tl) scintillation detectors. The in vivo precision of the technique, as estimated by three repeated measurements of 15 rats is 6 percent, for a radiation dose equivalent of 60 mSv. The radiation dose per measurement is sufficiently low to enable serial measurements on the same animal. MCNP-4A Monte Carlo transport code was utilized to calculate thermal neutron flux correction factors to account for differences in size and shape of the rats and calibration phantoms. Good agrement was observed in comparing body nitrogen assessment by prompt-(gamma) neutron activation and chemical carcass analysis.

A comparison is made between the techniques of Prompt Gamma Neutron Capture Analysis and Nuclear Resonance Absorption (NRA) of gamma-rays as applied to in-vivo determination of body elemental content in humans. Special emphasis is given to the quantification of body protein via the measurement of total body nitrogen. A proposed facility for in-vivo determination of nitrogen using the NRA method is described. The advantages and disadvantages of each method are listed and discussed.

I describe a new generation of experiments studying the weak interaction between nucleons. Measurements of the effect of this interaction are few in number and the significance of the observed effects are generally small. It is well known that the weak interaction violates parity. This was first experimentally established by C. S. Wu through measurement of an asymmetry of electrons emitted in the beta-decay of polarized ⁶⁰Co. The measured asymmetry was large because beta decay is a weak interaction process. For a process in which the strong interaction can contribute, we expect much smaller asymmetries, or order 10^-7.

Neutron resonance spectra in intermediate and heavy nuclei appear simply and regularly in the light of superfine structure (SFS). It is found that the relative shifts between the SFSs in groups of nuclei have definite values and are closely related to the symmetry properties of the compound systems if we treat the neutron zero energy as a common reference point. A possibility of time-based understanding of the SFSs is discussed. It is pointed out that the measurements on parity violation in neutron resonances are important to look into the nature of the SFS.

Experimental data on neutron induced spontaneously fissioning isomers are considered. To populate isometric states of heavy nuclei in the uranium-berkelium region (n, n') and (n, gamma) reactions on actinide targets were used; pulsed beams of fast neutrons were provided by the electrostatic accelerator facilities; time analysis was applied to fission fragment spectra to distinguish between prompt and delayed events and coincidence measurements to observe for the first time light charged particle accompanied fission from isomeric state were accomplished. Measured values for fission isomer yields and obtained shapes for excitation functions find their explanation beyond the scope of the nuclear shape isomerism hypothesis, the latter being the in-thing for nearly 30 years. To tell tow points apart the hints to quasiparticle, not the shape, nature of fission isomers are discussed.

We have designed and constructed a high-performance neutron polarimeter for measurements at intermediate energies, E_p equals 200-400 MeV. The current version of the neutron polarimeter, called NPOL2, consists of 6 planes of the 2D position-sensitive neutron detectors with a size of 1 m by 1 m by 0.1 m. Its effective left-right analyzing power has been calibrated at 141, 191, 291 and 387 MeV. The figure-of- merit value defined is about 4 X 10^-4, irrespective of neutron energies.

A procedure is presented by which the proton-neutron symmetry breaking in first excited 2⁺ stats of even- even nuclei can be estimated. The procedure involves calculations within the proton-neutron version of the interacting Boson model, and comparison with experimental data. The results of applying this procedure to even-even nuclei with 44 < Z < 80 are presented. In many cases, the amount of symmetry breaking, expressed in terms of the F-spin quantum number, is of the order to 10 percent.

MCNP is a general purpose 3D time-dependent neutron, photon, and electron transport code.It is highly portable and user- oriented, and backed by stringent software quality assurance practices and extensive experimental benchmarks. The cross section database is based upon the best evaluations available. MCNP incorporates state-of-the-art analog and adaptive Monte Carlo techniques. The code is documented in a 600 page manual which is augmented by numerous Los Alamos technical reports which detail various aspects of the code. MCNP represents over a megahour of development and refinement over the past 50 year and an ongoing commitment to excellence.

Monte Carlo methods for analyzing radiation transport are well established, although the ease of use can vary markedly between different software packages. This paper briefly reviews the present status of the MCBEND Monte Carlo code and how it has been developed to satisfy the requirements of the non-specialist user. It then discusses some of the ways in which MCBEND can be used to increase understanding of the performance of radiation systems by conducting virtual experiments.

An overview of Oak Ridge National Laboratory's (ORNL) 3D neural particle transport code, TORT, is presented. Special features of the code that make it invaluable for large applications are summarized for the prospective user. Advanced capabilities currently under development and installation in the production release of TORT are discussed in some detail. These include: multitasking on Cray platforms running the UNICOS operating system; adjacent-cell preconditioning acceleration scheme; and graphics codes for displaying computed quantities such as the flux. Further developments for TORT and its companion codes to enhance its present capabilities, as well as expand its range of applications will be discussed. Speculation on the next generation of neutral particle transport codes at ORNL, especially regarding unstructured grids and high order spatial approximations, will also be mentioned.

The capabilities of a program named GNSR which is designed to calculate response functions and efficiencies of various gas-filled neutron detectors, are described. Included are recoil detectors filled with H₂. ³He, ⁴He, and mixtures with commonly used quenching gases. A cylindrical detector geometry is assumed. The program calculates the detector response in various neutron beam configurations: point source, isotropic source, parallel beam, extended cylindrical source, and a plane rectangular localized source. The reliability of the code was tested with measurements in various quasi-monoenergetic neutron beams as also reported.

Changes introduced into the ANISN code in order to make available an option for solving neutron-transport problems with volumetric sources that may depend on space and angle are described. The usefulness of this new capability is illustrated for the case of a slab subject to a monodirectional external flux.

The kinetic energy of neutrons produced in the fusion plasma is converted in reactor components into heat. The calculated amount of nuclear heat and its distribution in the components has to be validated quantitatively. Well- established methods for heating measurements are based on the use of thermoluminescent detectors or micro-ionization chambers. These should be complemented by independent techniques of - if possible - higher sensitivity. Two such techniques have been developed at the Technical University of Dresden using scintillation detectors and a silicon microsensor. Both methods are described and results of methodical investigations are presented.

A solid iron slab with a vertical gap at different positions are irradiated with 14 MeV neutrons. The spectral neutron and photon fluxes penetrating and leaking the assemblies are measured and compared with Monte Carlo calculations for benchmarking calculational tools of fusion reactor design.

Certain experiments involving irradiation of samples in nuclear reactors, specially those for studies in biology, chemistry, physics and materials science where alteration of properties are to be induced by fast neutrons, require the knowledge of the neutron spectrum, particularly in order to calculate the doses absorbed by the samples on the basis of averaged kerma factors. Aiming at a general characterization of the fast neutron spectrum at the irradiation facilities of the Portuguese Research Reactor - a 1 MW pool type reactor, with fuel elements of MTR shape -, in a way which does not depend on the core configuration and on the irradiation conditions, the following simplified calculation was carried out: 1) the face of the reactor core has been considered as being a uniform plane source of fission neutrons, and 2) using the MCNP-4A code, the spectrum of fast neutrons has been calculated at several distances along the central axis normal to the source, considering various combinations of diffusing media of various thickness.

Detailed numerical prototypes are essential to the design of efficient and cost-effective neutron and gamma-ray imaging systems. We have exploited the unique capabilities of an LLNL-developed radiation transport code (COG) to develop code modules capable of simulating the performance of neutron and gamma-ray imaging systems over a wide range of source energies. COG allows us to simulate complex, energy-, angle-, and time-dependent radiation sources, model 3D system geometries with 'real world' complexity, specify detailed elemental and isotopic distributions and predict the responses of various types of imaging detectors with full Monte Carlo accuracy. COG references detailed, evaluated nuclear interaction databases allowing users to account for multiple scattering, energy straggling, and secondary particle production phenomena which may significantly effect the performance of an imaging system but may be difficult or even impossible to estimate using simple analytical models. In this work we will present examples illustrating the use of these routines in the analysis of industrial radiographic systems for thick target inspection, non-intrusive luggage and cargo scanning systems, and international treaty verification.

The Haefely neutron generator (HNG) was installed at IGA in January 1984 with the sole purpose of conducting fusion related neutronics research. Due to a recent change in priorities, the generator is presently dedicated to teaching activities in the framework of advanced practicals for undergraduate physics students of EPFL. This paper describes numerical modeling of the HNG and emphasizes design aspects important in its current utilization.

The insertion of ports, for diagnostics and additional heating devices, in the vacuum vessel of the projected fusion test reactor ITER will decrease the shielding capacity of the structure, increasing the expected neutron flux in critical components such as the superconducting toroidal field coils. A study of the nuclear heating in the coils has been carried out with the Monte Carlo code MCNP. A complete 45 degree sector of the torus, including the blanket, the vessel and the prot has been modelled. In operating conditions, the ports will be closed with plugs filling the entire hole except for the apertures needed for the particular diagnostic. In the present work, we have considered the pre-collimator plug for the neutron camera diagnostic, a multicollimator for neutron flux detectors and spectrometers located outside the machine. The toroidal coil has been divided in a matrix type shape to calculate the nuclear heating in different layers. Results show that the increase of heating in the more exposed regions is substantial, due both to less shielding thickness and to neutron streaming. Since the bulk structure of the plug can be considered rather typical, the conclusion is that care must be taken in its design and that he possibility to reduce the size of the apertures in the shielding blanket should be taken into account.

The method of bucking evaluation, realized in the MOnte Carlo code MCS, is described. This method was applied for calculational analysis of well known light water experiments TRX-1 and TRX-2. The analysis of this comparison shows, that there is no coincidence between Monte Carlo calculations, obtained by different ways: the MCS calculations with given experimental bucklings; the MCS calculations with given bucklings evaluated on base of full core MCS direct simulations; the full core MCNP and MCS direct simulations; the MCNP and MCS calculations, where the results of cell calculations are corrected by the coefficients taking into the account the leakage from the core. Also the buckling values evaluated by full core MCS calculations have differed from experimental ones, especially in the case of TRX-1, when this difference has corresponded to 0.5 percent increase of K_eff value.

The errors of the neutron spectra calculated by a multi- group Monte Carlo code were investigated by comparing the spectra with those by a continuous-energy Monte Carlo code for the transport of approximately MeV neutrons through graphite piles. When mono-directional neutrons impinge on a 10 cm thick pile, the transmission spectrum calculated with the multi-group code shows a violent oscillation caused by the truncation approximation of the Legendre series for group-transfer cross sections. The oscillations diminish and the spectrum obtained by the multi-group code tends to agree well with one by the continuous-energy code, when one or several of the following conditions are satisfied; 1) the neutron source is isotropic, 2) the thickness of the piles is large enough, 3) the detector locates in the pile.

A new nodal Sn transport calculation code NSHEX has been developed to accurately estimate the 3D transport effect of fast reactors in hexagonal-z geometry. The intra-node distribution of source is expanded into quadratic polynomials, and the neutron angular distribution of node surface fluxes and intra-node fluxes is considered by Sn quadrature. In addition, some improvements have been made on the code in order to enhance the accuracy. The accuracy of the present method was validated through numerical analyses for a commercial scale fast reactor with the comparison of Monte Carlo method.

Accelerator based monoenergetic neutron sources are reviewed with special emphasis on the requirements for BNCT, on sources in the gap region, on neutron production up to 200 MeV, and on neutron radiography. In particular, the following neutron producing reactions are considered: ³H(p,n)³He, ⁶Li(p,n)⁶Be, ⁷Li(p,n)⁷Be, ⁹Be(p,n)⁹B, ¹⁰Be(p,n)¹⁰B, ¹⁰B(p,n)¹⁰C, ¹¹B(p,n)¹¹C, ¹²C(p,n)¹²N, ¹³C(p,n)¹³N, ¹⁴C(p,n)¹⁴N, ¹⁵N(p,n)¹⁵O, ³⁶Cl(p,n)³⁶Ar, ³⁹Ar(p,n)³⁹K, ²H(d,n)³He, ³H(d,n)⁴He, ⁷Li(d,n)⁸Be, ⁹Be(d,n)¹⁰B, ¹³C(d,n)¹⁴N, ¹⁵N(d,n)¹⁶O, ¹⁸O(d,n)¹⁹F, ¹H(t,n)³He, ¹H(⁷Li,n)⁷Be, ¹H(¹⁰Be,n)¹⁰B, ¹H(¹¹B,n)¹¹C, ¹H(¹³C,n)¹³N, ¹H(¹⁴C,n)¹⁴N, ¹H(¹⁵N,n)¹⁵O, ¹H(³⁶Cl,n)³⁶Ar, ¹H(³⁹Ar,n)³⁹K, ²H(t,n)⁴He, ²H(¹⁵N,n)¹⁶O.

A fission neutron source is operating in Rome at the ENEA Casaccia Research Center since 1971, consisting of a low power fast reactor named RSV-Tapiro. it is employed for a variety of experiments, including dosimetry, material testing, radiation protection and biology. In particular, application to experimental radiobiology includes studies of the biological action of neutrons in the whole-body irradiated animal, or in specialized systems in vivo or in vitro. For his purpose a vertical irradiation facility was originally constructed. Recently, a new horizontal irradiation facility has been designed to allow the exposure of larger samples or larger sample batches at one time. Dosimetry at the sample irradiation positions is routinely carried out by the conventional method of using two ion chambers. This physical dosimetry has recently been compared with the results of biological dosimetry based on the detection of chromosomal aberrations in peripheral blood human lymphocytes irradiated in vitro. A characterization of the radiation quality in the two configurations has been carried out by tissue equivalent proportional counter microdosimetry measurements. Information about the main characteristics of the reactor and the two irradiation facilities is provided and relevant results of the various measurements are summarized. Radiobiological results obtained using this neutron source are also briefly outlined.

To replace the ageing nuclear reactors used today for radio- isotope production, research and industrial applications, we propose to use a spallation neutron source, with neutron multiplication by fission. A 150 MeV cyclotron will be used to produce a proton beam hitting a molten Pb-Bi primary target producing about 0.8 primary neutron per incident proton. The primary spallation neutrons, moderated, strike a number of secondary targets containing a subcritical amount of ²³⁵U. Typical neutron fluxes at the targets location will be around 6.10¹³n/cm²s. Such a systems can be used, for example, to supply more than 50 percent of the world demand in ⁹⁹Mo, the most frequently used radio- isotope in nuclear medicine. The non-critical nature of the system makes it more acceptable for the public than a nuclear reactor and simplifies the licensing process. The foreseen investment and operational cost, including radiowaste and decommissioning, of such a system are more favorable than for nuclear reactors. In extra, the licensing procedure is considerably simplified and the time from planning to commissioning might be shorter than four years. This paper discusses the current status of the ADONIS- project and more particularly one of its applications: the optimization of this system as a dedicated ⁹⁹Mo production plant.

Compact high current RFQ linacs have proven to be reliable and robust accelerators for the generation of neutrons for a variety of applications. The ⁹Be(d,n)¹⁰B reaction is often preferred in such applications due to its high cross section for neutron production and the good thermal properties of metallic beryllium targets allowing for high beam power. Thick and thin beryllium target neutron and gamma spectra have been investigated in detail using the time-of-flight technique for deuteron energies of 0.9, 1.5 and 4.0 MeV. These energies were chosen to correspond to the energies of commercially available RFQ deuteron linacs. The results show characteristics of neutron and gamma ray production of importance for the applications of this neutron source.

A windowless deuterium gas target has been constructed for either monoenergetic or white neutron production with a 900 KeV deuteron accelerator. The target is capable of operation at 100 mbar target pressure, and can admit a beam of 5 mm transverse extent. This target is further being modified by the inclusion of an intermittent valve arrangement to reduce the flow rates in the higher pressure stages. This valve should allow operation at up to 1000 mbar with low duty factor beams.

A plasma focus is an rz-plasma pinch device that may preferentially produce either x-rays or neutrons. In its neutron-production mode, the neutron pulse duration is approximately 10¹² neutrons/pulse from the d(d,n)³He(dd) reaction. The equivalent neutron yield form the d(t,n)⁴He(dt) reaction is approximately 10¹⁴ neutrons/pulse. The smallest plasma focus devices have produced approximately 10⁷ neutrons/pulse from the (dd) reaction. Across this range of neutron yields, there are several applications for which the plasma focus seems to be well suited. With a total yield of approximately 10¹² neutrons/pulse from the (dd) reaction, static neutron radiography for inert objects has been demonstrated. For biological systems, (dt) reaction yields of about 10¹² to 10¹³ neutrons/pulse will produce acceptable neutron radiographs. Given the short duration of the neutron pulse and localized nature, (dt) neutron yields in the range of 10¹³ to 10¹⁴ neutrons/pulse provide a useful source for basic neutron physics studies. Finally, modest scaling of the plasma focus neutron yield to the range of 10¹⁵ to 10¹⁶ neutrons/pulse from the (dt) reaction may provide a method of producing some neutron-produced, medical radioisotopes.

Ultra Cold Neutrons (UCN) can be produced at spallation sources using a variety of techniques. To date the technique used has ben to Bragg scatter and Doppler shift cold neutrons into UCN from a moving crystal. This is particularly applicable to short-pulse spallation sources. We are presently constructing a UCN source at LANSCE using this method.In addition, large gains in UCN density should be possible using cryogenic UCN sources. Research is under way at Gatchina to demonstrate technical feasibility of a frozen deuterium source. If successful, a source of this type could be implemented at future spallation source, such as the long pulse source being planned at Los Alamos, with a UCN density that may be two orders of magnitude higher than that presently available at reactors.

For the nuclear design of a spectrum-controlled irradiation field in a research reactor, reaction rate distributions of gold, nickel and indium wires were measured in the Kyoto University Critical assembly in tight-pitch lattice cores of different moderator-to-fuel volume ratios with spectrum shifter regions which consist of beryllium, graphite and aluminum. Spectrum indices were obtained by measuring the cadmium ratio of gold wire and calculating the ratio of reaction rates between gold and nickel wires. The measured data were compared with the results of neutronics calculations which are employed in the neutronics design of the irradiation field.

We discuss the use and the installation of a post- acceleration beam-chopper as a necessary device in high resolution, large bandwidth spectrometry of fast neutrons based on time-of-flight measurements around a medium energy cyclotron.

Low energy neutrons are easily thermalized and are useful in a number of applications including radiography and neutron scattering. A consideration of the early literature and recent experimental results suggests that the reaction ⁹Be(d,n)¹⁰B can provide a useful source of neutrons in the region of zero to a few hundred KeV that fulfill this requirement. The characteristics of the reaction are describe din this context and some of the limitations on intensity are considered. The advantages of a low energy source in terms of thermalization have been investigated using a MOnte Carlo simulation and it is proposed that a high current deuteron accelerator with an energy in the region of 1.1 to 1.5 MeV, used with a beryllium target provides a neutron source that merits serious consideration.

Methods for detecting neutrons, thermal to high energy, and for determining neutron spectra are reviewed, with emphasis on work published in the years 1992 to 1996.

A sixteen-segment NaI(Tl) multiplicity gamma ray detector is used at the Rensselaer Polytechnic Institute Gaerttner LINAC Laboratory for neutron cross section measurements. This detector consists of an annulus of NaI(Tl) divided into two sets of 8 pie shaped segments, each segment optically isolate and viewed by a photomultiplier. The neutron beam passes along the axis of the detector and impinges upon a sample placed in the center. Time-of-flight data are taken as a function of the number of sections which detect a gamma and which is defined as the detected multiplicity. This detector can simultaneously acquire neutron scattering, capture and fission data by placing suitable limits on the total detected gamma ray energy deposited in the detector. Scattering and capture measurements have been performed on samples of holmium, erbium, and tungsten and experimental results are presented. The experimental multiplicity for capture is analyzed by assuming the single particle model, stochastically calculating the gamma ray cascades from neutron capture, and transporting each gamma ray into the detector using the Monte Carlo method. The detection efficiency for neutron capture is over 90 percent and is relatively insensitive to different isotopes of the same element or different spins of the compound nuclear resonances. A status report on experimental and analytical activities at the Laboratory is presented.

With our work we want to demonstrate that position-sensitive silicon detectors, in combination with gadolinium foil converters, can successfully be used for the detection of thermal neutrons. We set up and tested different detector systems. Starting with a single diode and conventional electronics we determined the efficiency of the Si-Gd- Detector for different neutron wavelengths for both natural Gd and ¹⁵⁷Gd as converter. As a next step, we constructed a 16 channel system consisting of a 1D Si-strip detector with 1 mm pitch of strips and integrated readout electronics with self-triggering capability. Encouraged by the results obtained, we started development of a 2D system. We increased the position resolution, as well as the number of channels, and produced a 2D device based on a pad detector with 300 300micrometers ² pads.

Commercial dynamic Random Access Memory (dRAM) chips can offer a simple approach for the detection of neutrons. Indeed, ionizing radiation induced errors in the stored data of these silicon based memory deices provided the foundation for a new type of generic radiation sensor with the potential of high spatial resolution. Experiments were carried out and a MOnte Carlo program has been used to examine the response of the dynamic memory structures by which the sensitivity of dRAMs to neutrons can be increased. The aim of this paper is to present recent results and to discuss the capabilities of such a neutron detecting deice.

High accuracy imaging of thermal neutrons has been demonstrated with detectors having solid converters coated with secondary electron emitters and coupled to gaseous electron multipliers. The present work concentrates on study of the neutron detection efficiency of detectors equipped with composite CsI coated ⁶Li neutron converters. Monte Carlo simulations of the neutron conversion and the charged particle and electron emission processes permit to optimize the converter parameters. The experimental study of the detection efficiency is described in detail and the results are compared with the theoretical predictions.

Capillary neutron optics improve the capabilities of neutron beam techniques such as neutron depth profiling and prompt gamma activation analysis. Millions of glass capillaries are configured to capture and guide low-energy neutrons by grazing total reflection from the smooth inner surface of the hollow channels. By precise orientation of the capillaries, beams of neutrons are readily collimated with good angular control or can be finely focused - as required by the application. In addition, the optics can improve the signal-to-noise ratio by diverting a neutron beam to a convenient off-axis direction, thereby circumventing interferences from gamma rays and fast neutrons characteristic of simple aperture collimation. The focused intensity of neutrons obtained in an area of 0.03 mm² may be increased up to a hundred times over that previously available for NDP or PGAA techniques. Furthermore, the spatial resolution can be improved by up to 100 times. Consequently, small samples, or small volumes within larger samples, may be better and more rapidly investigated with neutron probe techniques. We report on developments in the application of capillary neutron optics.

A set-up based on a 16-section NaI(Tl) scintillation detector has been created at the 500 m flight-path of the IBR-30 booster in Dubna. The total volume of the detector is 36 liters, and the geometric efficiency is 80 percent. This detector measures gamma-cascades from resolved neutron resonances to obtain gamma-multiplicity spectra from neutron capture events. The ¹⁰B loaded converter, positioned inside the detector, makes it possible to detect also neutron scattering events by measuring monoenergetic gamma- quanta from the ¹⁰B(n, (alpha) (gamma) ) reaction. Thus, the same detector simultaneously registers neutron capture and neutron scattering events in the target under identical conditions. This permits spins, neutron and radiative widths, and neutron strength functions of resonances, to be determined. Applications of the detector for gamma- multiplicity studies are presented in the example of several Hf, Sn and Sm isotopes.

Neutron detection efficiencies for a 4 (pi) plastic scintillator detector were calculated with the Monte Carlo code of Cecil et al for various threshold of 2.5, 5, 7.5, 10 and 20 MeV_ee, assuming an isotropic distribution of neutrons emerging form the center of the target. A good agreement between the data obtained from the reaction (pi) ⁺⁴He yields pppn with 120 MeV pions and efficiency calculations has been found. Proton reaction losses for 15, 20 and 35 cm thick plastic detectors were calculated using the GEANT code for proton energies up to 150 MeV. Reaction losses were extracted from the simulated spectra. Our correction curve for proton reaction losses gives slightly higher values than the data of Measday and Serre and the calculation of Janni.

In multi-element neutron activation analysis with side nuclide half-life and concentration ranges, the experimental conditions have to be optimized in order to keep the count rates within certain limits, to avoid measurement distortions. By the use for a special loss-free counting technique which compensates the count losses due to high count rates and dead-times, the measurement time can be reduced drastically, permitting the measurement of short- lived and medium-lived nuclides in the presence of long- lived nuclides in spite of the possible high count rate immediately after activation, without matrix interference and short-lived nuclide information loss. In order to check the linearity and the reliability of the loss-free counting technique, several measurements have been performed by adding known amounts of a certain element to similar samples and by verifying the concentration values of standards at various count rates and dead-times, changing their activation level. The capability of short-time activation analysis increases the analytical throughput permitting more customers even far from the reactor facility to analyze samples by simply sending them to the reactor for irradiation and measurement and getting the data back for evaluation. Thus the application range can be broadened.

The need for studies of the activation of materials in the neutron fluxes encountered in D-T fusion devices has long been recognized as a key area of fusion technology programs. In this paper a comparison between experimental and numerical results of the induced radioactivity in a low activation material, the SiC/SiC fiber ceramic composite, is presented. The SiC/SiC fiber ceramic composite is a candidate material for the DEMO phase of a commercial fusion power station. This material was exposed to the neutron flux produced by a D-T neutron generator and the induced activity was carefully measured using the Low Background Counting Facility in the Gran Sasso laboratory. The neutron flux and spectra at the irradiation position have been determined using foil activation and numerical calculations with the Monte Carlo Code MCNP. These quantities have been used as input values for the activation code FISPACT and the cross section library EAF 4.1, to calculate the induced activity. The material composition and its impurities have been taken from existing data. The results of the comparison are discussed, focusing on the validation of the cross sections and the activity coming from the impurities.

After the Chernobyl catastrophe among the problems of current concern a question arose of detection of 'hot' particles formed from plutonium alloys with carbon, nitrogen, silicon, etc. For this purpose, the instruments are needed, which would be able to detect not only alpha- particles and low energy gamma-radiation, but also neutrons and high energy gamma-quanta from ((alpha) , n(gamma) ) - reactions. At present for each kind of radiation detectors of different types are used. A general drawback of all these instruments is their narrow dynamic range of dose rates and energies, and especially impossibility to registrate n-flux in condition large background activity gamma-rays nuclei, which makes each of them applicable only under certain specific conditions. For detection of 'hot' particles, oxide and semiconductor scintillators were used, which contained elements with large capture cross section for thermal neutrons. In this paper we try to determine possibilities and limitations of solid-state neutron detectors based on CdS(Te), ZnSe(Te), CdWO₄ (CWO), Gd₂SiO₅ (GSO) scintillators developed and produced by the Science and Technology Center for Radiation Instruments of the Institute for Single Crystals. The instruments developed by Center are based preferable on a very promising system 'scintillator- photodiode-preamplifier' matched with modern computer data processing techniques.

A review of methods currently used to unfold particle spectra form measured pulse height distributions or other detector readings is given. Algorithms based on the least- squares principle and the principle of maximum entropy are reviewed. By the examples presented, it is shown that most of the detection systems reveal ill-conditioned problems.

In measurements of fast neutron spectra by a ³He semiconductor detector, the unfolding method is not usually required. The unfolding method based on principle of maximum likelihood that incorporates the Gaussian approximation of counting statistics is developed and implemented in the MLMHE3 numerical code for application in fast neutron spectrometry by ³He semiconductor detectors. The derived likelihood equations have been solved using method of the singular value decomposition of the response matrix. For this inverse problem, the detector responses were generated by the Monte Carlo technique.

The output of a detector system recording neutron-induced nuclear reactions consists of a set of multichannel spectra and of scaler/counter values. These data must be reduced - i.e.. corrected and combined - to produce a clean energy spectrum of the reaction cross-section with a covariance estimate suitable for evaluation. The reduction process may be broken down into a sequence of operations. We present a set of reduction operations implemented as commands on a UNIX system. Every operation reads spectra from a file and appends results as new spectra to the same file. The binary file format AGS used thereby records the spectra as named entities including a set of neutron energy values and a corresponding set of values with their correlated and uncorrelated uncertainties.

We have studied the peak shapes at 596 and 691 KeV resulting from fast neutron interactions inside germanium detectors. We have used neutrons from a ²⁵²Cf source, as well as from the ²⁸Si((mu) ^-, nv), and ²⁰⁹Bi((pi) ^-, xn) reactions to compare the peaks and to check for any dependence of peak shape on the incoming neutron energy. In our investigation, no dependence of these peak shapes on the neutron energy spectra has been observed. In a comparison of these peak shapes with other studies we found similar results to ours except for monoenergetic neutron irradiations from 1 to 8 MeV.

This paper deals with the evaluation of neutron spectra in a wide energy range by unfolding the responses of different passive detectors. The unfolding method is based on the BUNKI code which has been optimized for this application. The major goal of this investigation is to obtain a better evaluation of the neutron dose equivalent both in accelerator environments and cosmic ray fields. This unfolding method has been tested in the neutron field at the CERN SPS facility, which is characterized by a neutron spectrum similar to that of cosmic rays. Finally the same procedure has been applied to evaluate the neutron cosmic ray spectrum on the mountain of Matterhorn at an altitude of 3500 m. In particular the high energy region of the neutron spectrum has been obtained taking into account the difference between the proton and neutron cross sections of bismuth, as recently proposed by one of the authors on the basis of theoretical and experimental data re-examination.

Energy resolution and difference of maximum energy of Compton electrons from half-height of Compton edge was measured experimentally for cyclohexane and deuterated- cycloheane (C₆D₁₂) scintillators for monoenergetic gamma-rays using gamma-gamma coincidence technique. For 0.5- 1.2 MeV gamma rays, the energy resolution of C₆D₁₂ and C₆H₁₂ scintillators vary from 26.6 to 14.0 percent and 31.1 to 15.7 percent respectively. The corresponding energy difference between the maximum Compton electron energy and the half height of the Compton edge for C₆D₁₂ and C₆H₁₂ scintillators vary form 15.3 to 6.3 percent and 14.0 to 7.33 percent respectively. Energy resolution of C₆D₁₂ and C₆H₁₂ scintillators is 14-31 percent and 18-31 percent poorer than that of the identical C₆D₆ and C₆H₆ scintillators.

The response of an NE213 scintillator has been determined for photons induced in the detector assembly by neutrons with energies between 1 MeV and 16 MeV. Coupled MCNP4A/EGS4 Monte Carlo simulations were compared and adjusted to experimental data. The complete set of response matrices for neutrons, photons and neutron induced photons was then used to determined the spectral neutron and photon fluence of ²³⁹Pu/¹³C and ²⁴¹Am/⁹Be sources.

Neutron response functions of two commercially available cylindrical ⁴He proportional counters have been measured in the neutron energy range from 500 to 8000 KeV. Experimental results are compared with calculated response functions, determined by Monte Carlo integration. Optimum operating conditions are determined and the detectors are calibrated both for energy and resolution. Energy and resolution calibration functions are expressed in terms of the stopping power of the charged particles in the detector. This allows the implementation of these functions in the Monte Carlo code GNSR, which has been developed of the calculation of response functions and efficiencies of gas- filled neutron detectors.

ICRP 60 publication will increase the requirements on an individual neutron dosimeter, particularly as far as the lowest limit of detection is concerned. Bubble damage detectors are at the moment probably the only devices capable to fulfil these requirements. Basic dosimetric characteristics of both these detectors have been studied. Results obtained are presented, analyzed and discussed.

Neutrons contribute to the exposure level at subsonic flight altitudes roughly by one half. The data on their spectrum differ rather substantially. A useful tool to improve the interpretation represents a calibration in properly formed reference fields. Another possibility is to compare the results with the reading of a reference instrument. Both approaches, together with analysis of available spectral data are presented in this contribution, an estimation of neutron exposure of aircrews at civil flight levels is given.