定价： 506元 / 折扣价： 431 元 加购物车

A lithium fluoride (LiF)-based photo-fluorescent film dosimetry system provides a means of determining absorbed dose to materials by the photo-stimulated emission of wavelengths longer than that of the stimulation wavelength. The absorbed dose is obtained from the amount of the light emission. Imperfections within the ionic lattice of alkali-halide compounds such as LiF act as traps for electrons and electron holes (positively charged negative-ion vacancies). These imperfections are known as color centers because of the part they play in the compound's ability to absorb and then release energy in the form of visible-light photons. Like an atom, these color centers have discrete, allowed energy levels, and electrons can be removed from these sites when energy of the appropriate wavelength and intensity is transferred to the material. The resulting fluorescence spectra contain discrete peaks that can cover a range of wavelengths, depending upon the type of alkali-halide (8). An example of fluorescence spectra from a LiF-based dosimeter is provided in Fig. 1. A system of optical filters within a light-detecting instrument (that is, fluorimeter) can be used to block all but a narrow range of wavelengths that are desired for use. Theories on how color centers are formed, how luminescence mechanisms work, and their application in dosimetry are found in Refs (8-13). For characterization studies on specific photo-fluorescent dosimeters see Refs (1-7) and (14-19).In the application of a specific dosimetry system, absorbed dose is determined by use of an experimentally-derived calibration curve. The calibration curve for the photo-fluorescent dosimeter is the functional relationship between ΔEf and D, and is determined by measuring the net fluorescence of sets of dosimeters irradiated to known absorbed doses. These absorbed doses span the range of utilization of the system.Photo-fluorescent dosimetry systems require calibration traceable to national standards. See ISO/ASTM Guide .The absorbed dose is usually specified relative to water. Absorbed dose in other materials may be determined by applying the conversion factors discussed in ISO/ASTM Guide .During calibration and use, possible effects of influence quantities such as temperature, light exposure, post-irradiation stabilization of signal, and absorbed-dose rate need to be taken into account.Photo-fluorescent dosimeters are sensitive to light, especially during irradiation and post-irradiation stabilization (7). Some color centers are sensitive to the UV and blue regions of the spectrum, while other centers are only sensitive to the UV. Therefore, they need to be packaged in appropriate light-tight packaging shortly after manufacture, and during use they need to be packaged or the appropriate filters placed over room lighting. Filtering the light fixtures involved during irradiation may be required for irradiations using low-energy X-rays or electrons where unpackaged dosimeters are used.The signal from photo-fluorescent dosimeters either increases or decreases with time following irradiation, depending on the color center utilized (19). This stabilization process, which can last from hours to days depending on storage temperature (and dose for some color centers) can be accelerated and stabilized by heat treating the dosimeters after irradiation and before readout (see 9.2).Note—Also shown are transmission curves for green and red emission filters.FIG. 1 Excitation Spectrum and Resulting Fluorescence Spectrum from the Sunna LiF-based Film Dosimeter1.1 This practice covers the handling, testing, and procedure for using a lithium fluoride (LiF)-based photo-fluorescent film dosimetry system to measure absorbed dose (relative to water) in materials irradiated by photons or electrons. Other alkali halides that may also exhibit photofluorescence (for example, NaCl, NaF, and KCl) are not covered in this practice. Although various alkali halides have been used for dosimetry for years utilizing thermoluminescence, the use of photoluminescence is relatively new.1.2 This practice applies to photo-fluorescent film dosimeters (referred hereafter as photo-fluorescent dosimeters) that can be used within part or all of the following ranges:1.2.1 Absorbed dose range of 5 × 10-2 to 3 × 102 kGy (1-3).1.2.2 Absorbed dose rate range of 0.3 to 2 × 104 Gy/s (2-5)).1.2.3 Radiation energy range for photons of 0.05 to 10 MeV (2).1.2.4 Radiation energy range for electrons of 0.1 to 10 MeV (2).1.2.5 Radiation temperature range of -20 to +60°C (6,7).1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

定价： 0元 / 折扣价： 0 元

定价： 646元 / 折扣价： 550 元 加购物车

4.1 The Fricke dosimetry system provides a reliable means for measurement of absorbed dose to water, based on a process of oxidation of ferrous ions to ferric ions in acidic aqueous solution by ionizing radiation (ICRU 80, PIRS-0815, (4)). In situations not requiring traceability to national standards, this system can be used for absolute determination of absorbed dose without calibration, as the radiation chemical yield of ferric ions is well characterized (see Appendix X3).4.2 The dosimeter is an air-saturated solution of ferrous sulfate or ferrous ammonium sulfate that indicates absorbed dose by an increase in optical absorbance at a specified wavelength. A temperature-controlled calibrated spectrophotometer is used to measure the absorbance.1.1 This practice covers the procedures for preparation, testing, and using the acidic aqueous ferrous ammonium sulfate solution dosimetry system to measure absorbed dose to water when exposed to ionizing radiation. The system consists of a dosimeter and appropriate analytical instrumentation. The system will be referred to as the Fricke dosimetry system. The Fricke dosimetry system may be used as either a reference standard dosimetry system or a routine dosimetry system.1.2 This practice is one of a set of standards that provides recommendations for properly implementing dosimetry in radiation processing, and describes a means of achieving compliance with the requirements of ISO/ASTM Practice 52628 for the Fricke dosimetry system. It is intended to be read in conjunction with ISO/ASTM Practice 52628.1.3 The practice describes the spectrophotometric analysis procedures for the Fricke dosimetry system.1.4 This practice applies only to gamma radiation, X-radiation (bremsstrahlung), and high-energy electrons.1.5 This practice applies provided the following are satisfied:1.5.1 The absorbed dose range shall be from 20 Gy to 400 Gy (1).21.5.2 The absorbed dose rate does not exceed 106 Gy·s−1 (2).1.5.3 For radioisotope gamma sources, the initial photon energy is greater than 0.6 MeV. For X-radiation (bremsstrahlung), the initial energy of the electrons used to produce the photons is equal to or greater than 2 MeV. For electron beams, the initial electron energy is greater than 8 MeV.NOTE 1: The lower energy limits given are appropriate for a cylindrical dosimeter ampoule of 12 mm diameter. Corrections for displacement effects and dose gradient across the ampoule may be required for electron beams (3). The Fricke dosimetry system may be used at lower energies by employing thinner (in the beam direction) dosimeter containers (see ICRU Report 35).1.5.4 The irradiation temperature of the dosimeter should be within the range of 10 °C to 60 °C.1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

定价： 702元 / 折扣价： 597 元 加购物车

Radiation processing of articles in both commercial and research applications may be carried out for a number of purposes. These include, for example, sterilization of health care products, reduction of the microbial populations in foods and modification of polymers. The radiations used may be accelerated electrons, gamma-radiation from radionuclide sources such as cobalt-60, or X-radiation.To demonstrate control of the radiation process, the absorbed dose must be measured using a dosimetry system, the calibration of which, is traceable to appropriate national or international standards. The radiation-induced change in the dosimeter is evaluated and related to absorbed dose through calibration. Dose measurements required for particular processes are described in other standards referenced in this practice.1.1 This practice describes the basic requirements that apply when making absorbed dose measurements in accordance with the ASTM E10.01 series of dosimetry standards. In addition, it provides guidance on the selection of dosimetry systems and directs the user to other standards that provide specific information on individual dosimetry systems, calibration methods, uncertainty estimation and radiation processing applications.1.2 This practice applies to dosimetry for radiation processing applications using electrons or photons (gamma- or X-radiation).1.3 This practice addresses the minimum requirements of a measurement management system, but does not include general quality system requirements.1.4 This practice does not address personnel dosimetry or medical dosimetry.1.5 This practice does not apply to primary standard dosimetry systems.1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

定价： 0元 / 折扣价： 0 元

定价： 590元 / 折扣价： 502 元 加购物车

4.1 Integral Mode Dosimetry—As shown in 3.2, two different integral relationships can be established using proton-recoil emulsion data. These two integral reactions can be obtained with roughly an order of magnitude reduction in scanning effort. Consequently, this integral mode is an important complementary alternative to the customary differential mode of NRE spectrometry. The integral mode can be applied over extended spatial regions, for example, perhaps up to as many as ten in-situ locations can be covered for the same scanning effort that is expended for a single differential measurement. Hence the integral mode is especially advantageous for dosimetry applications which require extensive spatial mapping, such as exist in Light Water Reactor-Pressure Vessel (LWR-PV) benchmark fields (see Test Method E1005). In low power benchmark fields, NRE can be used as integral dosimeters in a manner similar to RM, solid state track recorders (SSTR) and helium accumulation monitors (HAFM) neutron dosimeters (see Test Methods E854 and E910). In addition to spatial mapping advantages of these other dosimetry methods, NRE offer fine spatial resolution and can therefore be used in-situ for fine structure measurements. In integral mode scanning, both absolute reaction rates, that is I(ET) and J(Emin), are determined simultaneously. Separate software codes need to be used to permit operation of a computer based interactive system in the integral mode (see Section 9). It should be noted that the integrals I(ET) and J(Emin) possess different units, namely proton-recoil tracks/MeV per hydrogen atom and proton-recoil tracks per hydrogen atom, respectively. 4.2 Applicability for Spectral Adjustment Codes—In the integral mode, NRE provide absolute integral reaction rates that can be used in neutron spectrum least squares adjustment codes (see Guide E944). In the past, such adjustment codes could not utilize NRE integral reaction rates because of the non-existence of NRE data. NRE integral reaction rates provide unique benchmark data for use in least squares spectral adjustment codes. The unique significance of NRE integral data arises from a number of attributes, which are described separately below. Thus, inclusion of NRE integral reaction rate data in the spectral adjustment calculations can result in a significant improvement in the determination of neutron spectra in low power benchmark fields. 4.3 The Neutron Scattering Cross Section of Hydrogen—Integral NRE reaction rates are based on the standard neutron scattering cross section of hydrogen. For fast neutron spectrometry and dosimetry applications, the accuracy of this (n,p) cross section over extended energy regions is essentially unmatched. A semi-empirical representation of the energy-dependence of the (n,p) cross section is given in Eq 13. where: E is in MeV and σnp(E) is in barns. This energy-dependent representation of the (n,p) cross section possesses an uncertainty of approximately 1 % at the (1σ) level (19). 4.4 Threshold Energy Definition—In contrast with all other fast neutron dosimetry cross sections, the threshold energy of the I and J integral reaction rates can be varied. NRE integral reaction threshold variability extends down to approximately 0.3 to 0.4 MeV, which is the lower limit of applicability of the NRE method. Threshold variation is readily accomplished by using different lower bounds of proton track length to analyze NRE proton-recoil track length distributions. Furthermore, these NRE thresholds are more accurately defined than the corresponding thresholds of all other fast neutron dosimetry cross sections. NRE therefore provide a response with an extremely sharp energy cutoff that is not only unmatched by other cross sections, but an energy threshold that is independent of the in-situ neutron spectrum. No other fast neutron dosimetry cross sections possess a threshold response with these significant attributes. The behavior of the I-integral and J-integral response for different threshold energies is shown in Figs. 2 and 3, respectively, in comparison to the threshold 237Np(n,f) reaction used in RM dosimetry. FIG. 2 Comparison of the I-Integral Response with the 237Np (n,f) Threshold Reaction FIG. 3 Comparison of the J-Integral Response for ET = 0.404, 0.484, 0.554 and 0.620 MeV with the 237Np (n,f) Threshold Reaction 4.5 Complimentary Energy Response—It is of interest to compare the differential energy responses available from these two integral relations. From Eq 4 and 11, one finds responses of the form σ(E)/ E and (1 –Emin/E)σ(E) for the I and J integral relations, respectively. These two responses are compared in Fig. 4 using a common cut-off of 0.5 MeV for both ET and Emin. Since these two responses are substantially different, simultaneous application of these two integral relations would be highly advantageous. As shown in Fig. 4, the energy response of the I and J integral reaction rates complement each other. The J-integral response increases with increasing neutron energy above the threshold value and therefore possesses an energy dependence qualitatively similar to most fast neutron dosimetry cross sections. However, significant quantitative differences exist. As discussed above, the J-integral response is more accurately defined in terms of both the energy-dependent cross section and threshold energy definition. The I-integral possesses a maximum value at the threshold energy and decreases rapidly from this maximum value as neutron energy increases above the threshold value. As can be seen in Fig. 4, the I-integral possesses a much more narrowly defined energy response than the J-integral. While the J-integral response is broadly distributed, most of the I-integral response is concentrated in the neutron energy just above threshold. As a consequence, the I-integral reaction rate data generally provides a more rigorous test of the ability of neutron transport calculations to describe the complex spatial and energy variations that exist in benchmark fields than does the J-integral data. This conclusion is supported by the calculation to experiment ratios (C/E) obtained from NRE experiments in the VENUS-1 LWR-PV benchmark field. For these VENUS-1 NRE experiments, the C/E values for the I integral possessed larger variation and deviated more widely from unity than the corresponding C/E values for the J-integral (20). FIG. 4 Energy Dependent Response for the Integral Reactions I(ET) and J(Emin) 1.1 Nuclear Research Emulsions (NRE) have a long and illustrious history of applications in the physical sciences, earth sciences and biological sciences (1, 2)2. In the physical sciences, NRE experiments have led to many fundamental discoveries in such diverse disciplines as nuclear physics, cosmic ray physics and high energy physics. In the applied physical sciences, NRE have been used in neutron physics experiments in both fission and fusion reactor environments (3-6). Numerous NRE neutron experiments can be found in other applied disciplines, such as nuclear engineering, environmental monitoring and health physics. Given the breadth of NRE applications, there exist many textbooks and handbooks that provide considerable detail on the techniques used in the NRE method (1-4, 6). As a consequence, this practice will be restricted to the application of the NRE method for neutron measurements in reactor physics and nuclear engineering with particular emphasis on neutron dosimetry in benchmark fields (see Matrix E706). 1.2 NRE are passive detectors and provide time integrated reaction rates. As a consequence, NRE provide fluence measurements without the need for time-dependent corrections, such as arise with radiometric (RM) dosimeters (see Test Method E1005). NRE provide permanent records, so that optical microscopy observations can be carried out any time after exposure. If necessary, NRE measurements can be repeated at any time to examine questionable data or to obtain refined results. 1.3 Since NRE measurements are conducted with optical microscopes, high spatial resolution is afforded for fine structure experiments. The attribute of high spatial resolution can also be used to determine information on the angular anisotropy of the in-situ neutron field (4, 5, 7). It is not possible for active detectors to provide such data because of in-situ perturbations and finite-size effects (see Section 11). 1.4 The existence of hydrogen as a major constituent of NRE affords neutron detection through neutron scattering on hydrogen, that is, the well known (n,p) reaction. NRE measurements in low power reactor environments have been predominantly based on this (n,p) reaction. NRE have also been used to measure the 6Li (n,t) 4He and the 10B (n,α) 7Li reactions by including 6Li and 10B in glass specks near the mid-plane of the NRE (8, 9). Use of these two reactions does not provide the general advantages of the (n,p) reaction for neutron dosimetry in low power reactor environments (see Section 4). As a consequence, this standard will be restricted to the use of the (n,p) reaction for neutron dosimetry in low power reactor environments. 1.5 Limitations—The NRE method possesses four major limitations for applicability in low power reactor environments. 1.5.1 Gamma-Ray Sensitivity—Gamma-rays create a significant limitation for NRE measurements. Above a gamma-ray exposure of approximately 0.025 Gy, NRE can become fogged by gamma-ray induced electron events. At this level of gamma-ray exposure, neutron induced proton-recoil tracks can no longer be accurately measured. As a consequence, NRE experiments are limited to low power environments such as found in critical assemblies and benchmark fields. Moreover, applications are only possible in environments where the buildup of radioactivity, for example, fission products, is limited. 1.5.2 Low Energy Limit—In the measurement of track length for proton recoil events, track length decreases as proton-recoil energy decreases. Proton-recoil track length below approximately 3μm in NRE cannot be adequately measured with optical microscopy techniques. As proton-recoil track length decreases below approximately 3 μm, it becomes very difficult to measure track length accurately. This 3-μm track length limit corresponds to a low energy limit of applicability in the range of approximately 0.3 to 0.4 MeV for neutron induced proton-recoil measurements in NRE. 1.5.3 High-Energy Limits—As a consequence of finite-size limitations, fast-neutron spectrometry measurements are limited to ≤15 MeV. The limit for in-situ spectrometry in reactor environments is ≤8MeV. 1.5.4 Track Density Limit—The ability to measure proton recoil track length with optical microscopy techniques depends on track density. Above a certain track density, a maze or labyrinth of overlapping tracks is created, which precludes the use of optical microscopy techniques. For manual scanning, this limitation arises above approximately 104 tracks/cm2, whereas interactive computer-based scanning systems can extend this limit up to approximately 105 tracks/cm2. These limits correspond to neutron fluences of 106 − 10 7 cm−2, respectively. 1.6 Neutron Spectrometry (Differential Measurements)—For differential neutron spectrometry measurements in low-power reactor environments, NRE experiments can be conducted in two different modes. In the more general mode, NRE are irradiated in-situ in the low power reactor environment. This mode of NRE experiments is called the 4π mode, since the in-situ irradiation creates tracks in all directions (see 3.1.1). In special circumstances, where the direction of the neutron flux is known, NRE are oriented parallel to the direction of the neutron flux. In this orientation, one edge of the NRE faces the incident neutron flux, so that this measurement mode is called the end-on mode. Scanning of proton-recoil tracks is different for these two different modes. Subsequent data analysis is also different for these two modes (see 3.1.1 and 3.1.2). 1.7 Neutron Dosimetry (Integral Measurements)—NRE also afford integral neutron dosimetry through use of the (n,p) reaction in low power reactor environments. Two different types of (n,p) integral mode dosimetry reactions are possible, namely the I-integral (see 3.2.1) and the J-integral (see 3.2.2) (10, 11). Proton-recoil track scanning for these integral reactions is conducted in a different mode than scanning for differential neutron spectrometry (see 3.2). Integral mode data analysis is also different than the analysis required for differential neutron spectrometry (see 3.2). This practice will emphasize NRE (n,p) integral neutron dosimetry, because of the utility and advantages of integral mode measurements in low power benchmark fields. 1.8 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

定价： 646元 / 折扣价： 550 元 加购物车

定价： 590元 / 折扣价： 502 元 加购物车

定价： 0元 / 折扣价： 0 元 加购物车

定价： 590元 / 折扣价： 502 元 加购物车

AbstractThese methods cover general procedures for the calibration of radiation detectors and the analysis of radionuclides. For each individual radionuclide, one or more of these methods may apply. These methods are concerned only with specific radionuclide measurements. The chemical and physical properties of the radionuclides are beyond the scope of this standard. Among the measurement standards discussed are: the calibration and usage of germanium detectors, scintillation detector systems, scintillation detectors for simple and complex spectra, and counting methods such as beta particle counting, aluminum absorption curve, alpha particle counting, and liquid scintillation counting. For each of the methods, the scope, apparatus used, summary of methods, preparation of apparatus, calibration procedure, measurement of radionuclide, performance testing, sources of uncertainty, precautions and tests, and calculations are detailed.1.1 This guide covers general procedures for the calibration of radiation detectors and measurement for radiation metrology for reactor dosimetry. For any particular radionuclide, one or more of these methods may apply.1.2 These techniques are concerned only with specific radionuclide measurements. The chemical and physical properties of the radionuclides are not within the scope of this standard.1.3 E3376, Standard Practice for Calibration and Usage of Germanium Detectors in Radiation Metrology for Reactor Dosimetry, was previously in Guide E181 and is now found in Volume 12.02 of the Annual Book of ASTM Standards. The discussion herein is not a sufficient substitute for the full standard. This guide is specifically NOT to be used as a direct reference to Practice E3376. Only the standard listed provides sufficient information to serve as a reference.1.4 Additional information on the setup, calibration, and quality control for radiometric detectors and measurements is given in Guide C1402 and Practice D7282.1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

定价： 646元 / 折扣价： 550 元 加购物车

定价： 646元 / 折扣价： 550 元 加购物车

定价： 646元 / 折扣价： 550 元 加购物车

定价： 590元 / 折扣价： 502 元 加购物车

定价： 777元 / 折扣价： 661 元 加购物车