ISO 18557 presents guidelines for sampling strategies and characterization processes to assess the contamination of soils, buildings and infrastructures, prior to remediation and/or to check that the remediation objectives have been met (final release surveys). The principles presented need to be appropriately graded as regards the specific situations concerned (size, level of contamination?). It can be used in conjunction with each country's key documentation.
ISO 18557 deals with characterization in relation to site remediation. It applies to sites contaminated after normal operation of older nuclear facilities. It could also apply to site remediation after a major accident, and in this case the input data will be linked to the accident involved.
ISO 18557 complements existing standards, notably concerning sampling, sample preservation and their transport, treatment and laboratory measurements, but also those related to in situ chemical and radiological measurements. References in the Bibliography contain links to appropriate documentation and techniques as required by individual member countries.
ISO 18557 does not apply to the following issues: execution of clean-up works, sampling and characterization of waste (conditioned or unconditioned) or to waste packages.
It does not apply to groundwater characterization (saturated zone).
Given the case-by-case nature of site remediation and decommissioning, the principles and guidance communicated in ISO 18557 are intended as general guidance only, not prescriptive requirements.

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ISO 19581 specifies a screening test method to quantify rapidly the activity concentration of gamma-emitting radionuclides, such as 131I, 132Te, 134Cs and 137Cs, in solid or liquid test samples using gamma-ray spectrometry with lower resolution scintillation detectors as compared with the HPGe detectors (see IEC 61563).
This test method can be used for the measurement of any potentially contaminated environmental matrices (including soil), food and feed samples as well as industrial materials or products that have been properly conditioned. Sample preparation techniques used in the screening method are not specified in ISO 19581, since special sample preparation techniques other than simple machining (cutting, grinding, etc.) should not be required. Although the sampling procedure is of utmost importance in the case of the measurement of radioactivity in samples, it is out of scope of ISO 19581; other international standards for sampling procedures that can be used in combination with ISO 19581 are available (see References [1],[2],[3],[4],[5],[6]).
The test method applies to the measurement of gamma-emitting radionuclides such as 131I, 134Cs and 137Cs. Using sample sizes of 0,5 l to 1,0 l in a Marinelli beaker and a counting time of 5 min to 20 min, decision threshold of 10 Bq·kg−1 can be achievable using a commercially available scintillation spectrometer [e.g. thallium activated sodium iodine (NaI(Tl)) spectrometer 2" ϕ × 2" detector size, 7 % resolution (FWHM) at 662 keV, 30 mm lead shield thickness].
This test method also can be performed in a "makeshift" laboratory or even outside a testing laboratory on samples directly measured in the field where they were collected.
During a nuclear or radiological emergency, this test method enables a rapid measurement of the sample activity concentration of potentially contaminated samples to check against operational intervention levels (OILs) set up by decision makers that would trigger a predetermined emergency response to reduce existing radiation risks[12].
Due to the uncertainty associated with the results obtained with this test method, test samples requiring more accurate test results can be measured using high-purity germanium (HPGe) detectors gamma-ray spectrometry in a testing laboratory, following appropriate preparation of the test samples[7][8].
ISO 19581 does not contain criteria to establish the activity concentration of OILs.

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ISO 19361:2017 applies to liquid scintillation counters and requires the preparation of a scintillation source obtained by mixing the test sample and a scintillation cocktail. The test sample can be liquid (aqueous or organic), or solid (particles or filter or planchet).
ISO 19361:2017describes the conditions for measuring the activity of beta emitter radionuclides by liquid scintillation counting[14][15].
The choice of the test method using liquid scintillation counting involves the consideration of the potential presence of other beta emitter radionuclides in the test sample. In this case, a specific sample treatment by separation or extraction is implemented to isolate the radionuclide of interest in order to avoid any interference with other beta-, alpha- and gamma-emitting radionuclides during the counting phase.
ISO 19361:2017 is applicable to all types of liquid samples having an activity concentration ranging from a few Bq·l−1 to 106 Bq·l−1. For a liquid test sample, it is possible to dilute liquid test samples in order to obtain a solution having an activity compatible with the measuring instrument. For solid samples, the activity of the prepared scintillation source shall be compatible with the measuring instrument.
The measurement range is related to the test method used: nature of test portion, preparation of the scintillator - test portion mixture, measuring assembly as well as to the presence of the co-existing activities due to interfering radionuclides.
Test portion preparations (such as distillation for 3H measurement, or benzene synthesis for 14C measurement, etc.) are outside the scope of this document and are described in specific test methods using liquid scintillation[2][3][4][5][6][7][8][9].

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ISO 19226:2017 provides a procedure for the evaluation of irradiation data in the region between the reactor core and the inside surface of the containment vessel, through the pressure vessel and the reactor cavity, between the ends of active fuel assemblies, given the neutron source in the core.
NOTE       These irradiation data could be neutron fluence or displacements per atom (dpa), and Helium production.
The evaluation employs both neutron flux computations and measurement data from in-vessel and cavity dosimetry, as appropriate. This document applies to pressurized water reactors (PWRs), boiling water reactors (BWRs), and pressurized heavy water reactors (PHWRs).
ISO 19226:2017 also provides a procedure for evaluating neutron damage properties at the reactor pressure vessel and internal components of PWRs, BWRs, and PHWRs. Damage properties are focused on atomic displacement damage caused by direct displacements of atoms due to collisions with neutrons and indirect damage caused by gas production, both of which are strongly dependent on the neutron energy spectrum. Therefore, for a given neutron fluence and neutron energy spectrum, calculations of the total accumulated number of atomic displacements are important data to be used for reactor life management.

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This document describes radon-222 spot measurement methods. It gives indications for carrying out spot measurements, at the scale of a few minutes at a given place, of the radon activity concentration in open and confined atmospheres.
This measurement method is intended for rapid assessment of the radon activity concentration in the air. The result cannot be extrapolated to an annual estimate of the radon activity concentration. This type of measurement is therefore not applicable for assessment of the annual exposure or for determining whether or not to mitigate citizen exposures to radon or radon decay products.
The measurement method described is applicable to air samples with radon activity concentration greater than 50 Bq·m−3.
NOTE    For example, using an appropriate device, the radon activity concentration can be spot measured in the soil and at the interface of a material with the atmosphere (see also ISO 11665-7[8]).

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This document describes continuous measurement methods for radon-222. It gives indications for continuous measuring of the temporal variations of radon activity concentration in open or confined atmospheres.
This document is intended for assessing temporal changes in radon activity concentration in the environment, in public buildings, in homes and in work places, as a function of influence quantities such as ventilation and/or meteorological conditions.
The measurement method described is applicable to air samples with radon activity concentration greater than 5 Bq/m3.

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This document describes spot measurement methods for determining the activity concentration of short-lived radon-222 decay products in the air and for calculating the potential alpha energy concentration.
This document gives indications for performing a spot measurement of the potential alpha energy concentration, after sampling at a given place for several minutes, and the conditions of use for the measuring devices.
The measurement method described is applicable for a rapid assessment of the potential alpha energy concentration. The result obtained cannot be extrapolated to an annual estimate potential alpha energy concentration of short-lived radon-222 decay products. Thus, this type of measurement is not applicable for the assessment of annual exposure or for determining whether or not to mitigate citizen exposures to radon or radon decay products.
This measurement method is applicable to air samples with potential alpha energy concentration greater than 5 nJ/m3.
NOTE       This document does not address the potential contribution of radon-220 decay products.

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This document describes integrated measurement methods for short-lived radon‑222 decay products[4]. It gives indications for measuring the average potential alpha energy concentration of short‑lived radon-222 decay products in the air and the conditions of use for the measuring devices.
This document covers samples taken over periods varying from a few weeks to one year. This document is not applicable to systems with a maximum sampling duration of less than one week.
The measurement method described is applicable to air samples with potential alpha energy concentration of short-lived radon-222 decay products greater than 10 nJ/m3 and lower than 1 000 nJ/m3.
NOTE       For informative purposes only, this document also addresses the case of radon-220 decay products, given the similarity in behaviour of the radon isotopes 222 and 220.

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This document outlines guidance for measuring radon-222 activity concentration and the potential alpha energy concentration of its short-lived decay products in the air.
The measurement methods fall into three categories:
a)    spot measurement methods;
b)    continuous measurement methods;
c)    integrated measurement methods.
This document provides several methods commonly used for measuring radon-222 and its short-lived decay products in air.
This document also provides guidance on the determination of the inherent uncertainty linked to the measurement methods described in its different parts.

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ISO 11665-11:2016 describes radon-222 test methods for soil gas using passive and active in-situ sampling at depth comprised between surface and 2 m.
ISO 11665-11:2016 gives general requirements for the sampling techniques, either passive or active and grab or continuous, for in-situ radon-222 activity concentrations measurement in soil gas.
The radon-222 activity concentration in the soil can be measured by spot or continuous measurement methods (see ISO 11665‑1). In case of spot measurement methods (ISO 11665‑6), the soil gas sampling is active only. On the other hand, the continuous methods (ISO 11665‑5) are typically associated with passive soil gas sampling.
The measurement methods are applicable to all types of soil and are determined according to the end use of the measurement results (phenomenological observation, definition or verification of mitigation techniques, etc.) taking into account the expected level of the radon-222 activity concentration.
These measurement methods are applicable to soil gas samples with radon activity concentrations greater than 100 Bq/m3.
NOTE          This part of ISO 11665 is complementary with ISO 11665‑7 for characterization of the radon soil potential.

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ISO/TS 18090-1:2015 is directly applicable to pulsed X-radiation with pulse duration of 0,1 ms up to 10 s. This covers the whole range used in medical diagnostics at the time of publication. Some specifications may also be applicable for much shorter pulses; one example is the air kerma of one pulse. Such a pulse may be produced, e.g. by X-ray flash units or high-intensity femtosecond-lasers. Other specifications are not applicable for much shorter pulses; one example is the time-dependent behaviour of the air kerma rate. This may not be measurable for technical reasons as no suitable instrument is available, e.g. for pulses produced by a femtosecond-laser.
ISO/TS 18090-1:2015 specifies the characteristics of reference pulsed radiation for calibrating and testing radiation protection dosemeters and dose rate meters with respect to their response to pulsed radiation. The radiation characteristics includes the following:
a)    time-dependent behaviour of the air kerma rate of the pulse;
b)    time-dependent behaviour of the X-ray tube high voltage during the pulse;
c)    uniformity of the air kerma rate within a cross-sectional area of the radiation beam;
d)    air kerma of one radiation pulse;
e)    air kerma rate of the radiation pulse;
f)     repetition frequency.
ISO/TS 18090-1:2015 does not define new radiation qualities. Instead, it uses those radiation qualities specified in existing ISO and IEC standards. This part of ISO/TS 18090 gives the link between the parameters for pulsed radiation and the parameters for continuous radiation specifying the radiation qualities. It does not specify specific values or series of values for the pulsed radiation field but specifies only those limits for the relevant pulsed radiation parameters that are required for calibrating dosemeters and dose rate meters and for determining their response depending on the said parameters.
The pulse parameters with respect to the phantom-related quantities were determined using conversion coefficients according to ISO 4037 (all parts). This is possible as the radiation qualities specified in existing ISO and IEC standards are used.
A given reference pulsed X-ray facility is characterized by the parameter ranges over which the full specifications and requirements according to this part of ISO/TS 18090 are met. Therefore, not all reference pulsed X-ray facilities can produce pulses covering the same parameter ranges.

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ISO 16637:2016 specifies the minimum requirements for the design of professional programmes to monitor workers exposed to the risk of internal contamination via inhalation by the use of radionuclides as unsealed sources in nuclear medicine imaging and therapy departments. It establishes principles for the development of compatible goals and requirements for monitoring programmes and, when adequate, dose assessment. It presents procedures and assumptions for the risk analysis, for the monitoring programmes, and for the standardized interpretation of monitoring data.
ISO 16637:2016 addresses the following items:
a)    purposes of monitoring and monitoring programmes;
b)    description of the different categories of monitoring programmes;
c)    quantitative criteria for conducting monitoring programmes;
d)    suitable methods for monitoring and criteria for their selection;
e)    information that has to be collected for the design of a monitoring programme;
f)     general requirements for monitoring programmes (e.g. detection limits, tolerated uncertainties);
g)    frequencies of measurements;
h)    procedures for dose assessment based on reference levels for routine and special monitoring programmes;
i)     assumptions for the selection of dose-critical parameter values;
j)     criteria for determining the significance of individual monitoring results;
k)    interpretation of workplace monitoring results;
l)     uncertainties arising from dose assessments and interpretation of bioassays data;
m)   reporting/documentation;
n)    quality assurance.
ISO 16637:2016 does not address the following:
-      monitoring and internal dosimetry for the workers exposed to laboratory use of radionuclides such as radioimmunoassay techniques;
-      monitoring and internal dosimetry for the workers involved in the operation, maintenance, and servicing of PET cyclotrons;
-      detailed descriptions of measuring methods and techniques;
-      dosimetry for litigation cases;
-      modelling for the improvement of internal dosimetry;
-      the potential influence of medical treatment of the internal contamination;
-      the investigation of the causes or implications of an exposure;
-      dosimetry for ingestion exposures and for contaminated wounds.

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ISO 12799:2015 describes a procedure for measuring the nitrogen content of UO2, (U,Gd)O2, and (U,Pu)O2 pellets. Nitrogen in nuclear fuel may be present either as elemental nitrogen or chemically combined in the form of nitrogenous compounds. The technique described herein serves to determine the total content of nitrogen excluding those compounds whose decomposition temperature is above 2 200 °C (most notably Pu and U nitrides).

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ISO 22765:2016 describes the ceramographic procedure used to prepare sintered (U,Pu)O2 pellets for qualitative and quantitative examination of the pellet microstructure.
The examinations are performed before and after thermal treatment or chemical etching.
They allow
-      observation of any cracks, intra- and intergranular pores or inclusions, and
-      measurement of the grain size, porosity and plutonium homogeneity distribution.
The mean grain diameter is measured by one of the classic methods: counting (intercept method), comparison with standard grids or typical images, etc.[2] The measurement of individual grain sizes requires uniform development of the microstructure over the entire specimen.
The plutonium cluster and pore distribution and localization are generally analysed by automatic image analysis systems. The plutonium distribution is usually revealed by chemical etching but alpha-autoradiography can also be used. The first technique avoids the tendency for autoradiography to exaggerate the size of plutonium-rich clusters due to the distance the alpha particles travel away from the source.

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ISO 12800:2017 gives guidelines on the determination of the specific surface area of as-fabricated uranium dioxide powder by volumetric or gravimetric determination of the amount of nitrogen adsorbed on the powder, and can be applied to other similar materials, e.g. U3O8, UO2-PuO2 powders, and other bodies with similar surface areas, e.g. powder granules or green pellets, provided that the conditions described are fulfilled. Modifications using other adsorbing gases are included.
The method is relevant as long as the expected value is in the range between 1 m2/g and 10 m2/g.

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The scope of ISO 18417:2017 covers
-      iodine sorbents for nuclear power plants, nuclear facilities, research and other nuclear reactors,
-      iodine sorbents for laboratories, including nuclear medicine, and
-      iodine sorbents for sampling equipment on sample lines.
ISO 18417:2017 applies to iodine sorbents manufacturers and operators in order to measure the actual performance of these sorbents and their sorption capacity for radioiodine.
ISO 18417:2017 applies to granulated and crushed iodine sorbents based on activated charcoal (hereinafter referred to as "sorbents") used for trapping gaseous radioiodine and its compounds. This document establishes the method and conditions for defining sorption capacity index in a laboratory.

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ISO 16645:2016 is applicable to medical electron linear accelerators i.e. linear accelerators with nominal energies of the beam ranging from 4 MV to 30 MV, including particular installations such as robotic arm, helical intensity modulated radiotherapy devices and dedicated devices for intra operative radiotherapy (IORT) with electrons.
The cyclotrons and the synchrotrons used for hadrontherapy are not considered.
The radiation protection requirements and recommendations given in ISO 16645:2016 cover the aspects relating to regulations, shielding design goals and other design criteria, role of the manufacturers, of the radiation protection officer or qualified expert and interactions between stakeholders, radiations around a linear accelerator, shielding for conventional and special devices (including shielding materials and transmission values, calculations for various treatment room configurations, duct impact on radiation protection) and the radiological monitoring (measurements).

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ISO 16639:2017 provides guidelines and performance criteria for sampling airborne radioactive substances in the workplace. Emphasis is on health protection of workers in the indoor environment.
ISO 16639:2017 provides best practices and performance-based criteria for the use of air sampling devices and systems, including retrospective samplers and continuous air monitors. Specifically, this document covers air sampling program objectives, design of air sampling and monitoring programs to meet program objectives, methods for air sampling and monitoring in the workplace, and quality assurance to ensure system performance toward protecting workers against unnecessary inhalation exposures.
The primary purpose of the surveillance of airborne activity concentrations in the workplace is to evaluate and mitigate inhalation hazards to workers in facilities where these can become airborne. A comprehensive surveillance program can be used to
-      determine the effectiveness of administrative and engineering controls for confinement,
-      measure activity concentrations of radioactive substances,
-      alert workers to high activity concentrations in the air,
-      aid in estimating worker intakes when bioassay methods are unavailable,
-      determine signage or posting requirements for radiation protection, and
-      determine appropriate protective equipment and measures.
Air sampling techniques consist of two general approaches. The first approach is retrospective sampling, in which the air is sampled, the collection medium is removed and taken to a radiation detector system and analysed for radioactive substance, and the concentration results made available at a later time. In this context, the measured air concentrations are evaluated retrospectively. The second approach is continuous real-time air monitoring so that workers can be warned that a significant release of airborne radioactivity may have just occurred. In implementing an effective air sampling program, it is important to achieve a balance between the two general approaches. The specific balance depends on hazard level of the work and the characteristics of each facility.
A special component of the second approach which can apply, if properly implemented, is the preparation of continuous air monitoring instrumentation and protocols. This enables radiation protection monitoring of personnel that have been trained and fitted with personal protective equipment (PPE) that permit pre-planned, defined, extended stay time in elevated concentrations of airborne radioactive substances. Such approaches can occur either as part of a planned re-entry of a contaminated area following an accidental loss of containment for accident assessment and recovery, or part of a project which involves systematic or routine access to radioactive substances (e.g. preparing process material containing easily aerosolized components), or handling objects such as poorly characterized waste materials that may contain radioactive contaminants that could be aerosolized when handled during repackaging. In this special case, the role of continuous air monitoring is to provide an alert to health physics personnel that the air concentrations of concern have exceeded a threshold such that the planned level of protection afforded by PPE has been or could be exceeded. This level would typically be many 10's or 100's of times higher than the derived air concentration (DAC) established for unprotected workers. The mo

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ISO 12183:2016 describes an analytical method for the electrochemical assay of pure plutonium nitrate solutions of nuclear grade, with a total uncertainty not exceeding ±0,2 % at the confidence level of 0,95 for a single determination (coverage factor, K = 2). The method is suitable for aqueous solutions containing more than 0,5 g/L plutonium and test samples containing between 4 mg and 15 mg of plutonium. Application of this technique to solutions containing less than 0,5 g/L and test samples containing less than 4 mg of plutonium requires experimental demonstration by the user that applicable data quality objectives will be met.
For some applications, purification of test samples by anion exchange is required before measurement to remove interfering substances when present in significant amounts.

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ISO 21484:2017 describes a method for determining the Oxygen-to-Metal (O/M) ratio in mixed uranium-plutonium oxide (U,Pu)O2 ± X pellets. The parameters given in the following paragraphs are relevant for pellets within a range of O/M ratio corresponding to 1,98 to 2,01. The method described in the document is adapted, with regard to the parameters, if the expected values of O/M ratio are outside the range.

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ISO 19238:2014 provides criteria for quality assurance and quality control, evaluation of the performance, and the accreditation of biological dosimetry by cytogenetic service laboratories.
ISO 19238:2014 addresses
a)    the confidentiality of personal information, for the customer and the service laboratory,
b)    the laboratory safety requirements,
c)    the calibration sources and calibration dose ranges useful for establishing the reference dose-effect curves that contribute to the dose estimation from chromosome aberration frequency and the minimum resolvable doses,
d)    the scoring procedure for unstable chromosome aberrations used for biological dosimetry,
e)    the criteria for converting a measured aberration frequency into an estimate of absorbed dose,
f)     the reporting of results,
g)    the quality assurance and quality control,
h)    informative annexes containing sample instructions for customer, sample questionnaire, sample of report, fitting of the low dose-response curve by the method of maximum likelihood and calculating the error of dose estimate, odds ratio method for cases of suspected exposure to a low dose, and sample data sheet for recording aberrations.

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ISO 21613:2015 describes a method for determining chlorine and fluorine in mixed (U,Pu)O2 powders and sintered pellets. It is applicable for the analysis of samples containing 5 µg.g−1 to 50 µg.g−1 of chlorine and 2 µg.g−1 to 50 µg.g−1of fluorine.
For UO2 powder and sintered pellets, refer to ISO 22875.

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ISO 16424:2012 is applicable to the evaluation of the homogeneity of Gd distribution within gadolinium fuel blends, and the determination of the Gd2O3 content in sintered fuel pellets of Gd2O3+UO2 from 1 % to 10 %, by measurements of gadolinium (Gd) and uranium (U) elements using ICP-AES.
After performing measurements of Gd and U elements using ICP-AES, if statistical methodology is additionally applied, homogeneity of Gd distribution within a Gd fuel pellet lot can also be evaluated. However, ISO 16424:2012 covers the statistical methodology only on a limited basis.

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ISO 29661:2012 defines terms and fundamental concepts for the calibration of dosemeters and equipment used for the radiation protection dosimetry of external radiation -- in particular, for beta, neutron and photon radiation. It defines the measurement quantities for radiation protection dosemeters and doserate meters and gives recommendations for establishing these quantities. For individual monitoring, it covers whole body and extremity dosemeters (including those for the skin and the eye lens), and for area monitoring, portable and installed dosemeters. Guidelines are given for the calibration of dosemeters and doserate meters used for individual and area monitoring in reference radiation fields. Recommendations are made for the position of the reference point and the phantom to be used for personal dosemeters.
ISO 29661:2012 also deals with the determination of the response as a function of radiation quality and angle of radiation incidence.
ISO 29661:2012 is intended to be used by calibration laboratories and manufacturers.

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ISO 18589-3:2015 specifies the identification and the measurement of the activity in soils of a large number of gamma-emitting radionuclides using gamma spectrometry. This non-destructive method, applicable to large-volume samples (up to about 3 000 cm3), covers the determination in a single measurement of all the γ-emitters present for which the photon energy is between 5 keV and 3 MeV.
ISO 18589-3:2015 can be applied by test laboratories performing routine radioactivity measurements as a majority of gamma-emitting radionuclides is characterized by gamma-ray emission between 40 keV and 2 MeV.
The method can be implemented using a germanium or other type of detector with a resolution better than 5 keV.
ISO 18589-3:2015 is addressed to people responsible for determining gamma-emitting radionuclides activity present in soils for the purpose of radiation protection.

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ISO 18589-2:2015 specifies the general requirements, based on ISO 11074 and ISO/IEC 17025, for all steps in the planning (desk study and area reconnaissance) of the sampling and the preparation of samples for testing. It includes the selection of the sampling strategy, the outline of the sampling plan, the presentation of general sampling methods and equipment, as well as the methodology of the pre-treatment of samples adapted to the measurements of the activity of radionuclides in soil.
ISO 18589-2:2015 is addressed to the people responsible for determining the radioactivity present in soil for the purpose of radiation protection. It is applicable to soil from gardens, farmland, urban, or industrial sites, as well as soil not affected by human activities.
ISO 18589-2:2015 is applicable to all laboratories regardless of the number of personnel or the range of the testing performed. When a laboratory does not undertake one or more of the activities covered by this part of ISO 18589, such as planning, sampling, or testing, the corresponding requirements do not apply.

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ISO 17099:2014 addresses the following:
a)    confidentiality of personal information for the customer and the laboratory;
b)    laboratory safety requirements;
c)    radiation sources, dose rates, and ranges used for establishing the calibration reference dose-effect curves allowing the dose estimation from CBMN assay yields and the minimum resolvable dose;
d)    performance of blood collection, culturing, harvesting, and sample preparation for CBMN assay scoring;
e)    scoring criteria;
f)     conversion of micronucleus frequency in binucleated cells into an estimate of absorbed dose;
g)    reporting of results;
h)    quality assurance and quality control;
i)     informative annexes containing examples of a questionnaire, instructions for customers, a microscope scoring data sheet, a sample report and advice on strengths and limitations of current automated systems for automated micronucleus scoring.

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ISO 19017:2015 is applicable to gamma radiation measurements on radioactive waste.
Radioactive waste can be found in different forms and exhibit a wide range of characteristics, including the following:
-      raw or unconditioned waste, including process waste (filters, resins, control rods, scrap, etc.) and waste from dismantling or decommissioning;
-      conditioned waste in various forms and matrices (bitumen, cement, hydraulic binder, etc.);
-      very low level (VLLW), low level (LLW), intermediate level (ILW) and high level radioactive waste (HLW);
-      different package shapes: cylinders, cubes, parallelepipeds, etc.
Guidance is provided in respect of implementation, calibration, and quality control. The diversity of applications and system realizations (ranging from research to industrial systems, from very low level to high level radioactive waste, from small to large volume packages with different shapes, with different performance requirements and allowable measuring time) renders it impossible to provide specific guidance for all instances; the objective of this International Standard is, therefore, to establish a set of guiding principles. Ultimately, implementation is to be performed by suitably qualified and experienced persons and based on a thorough understanding of the influencing factors, contributing variables and performance requirements of the specific measurement application.
This International Standard assumes that the need for the provision of such a system will have been adequately considered and that its application and performance requirements will have been adequately defined through the use of a structured requirements capture process, such as data quality objectives (DQO).
It is noted that, while outside the scope of this International Standard, many of the principles, measurement methods, and recommended practices discussed here are also equally applicable to gamma measurements of items other than radioactive waste (e.g. bulk food, water, free-standing piles of materials) or to measurements made on radioactive materials contained within non-traditional packages (e.g. in transport containers).

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The following documents, in whole or in part, are normatively referenced in ISO 20785-3:2015 and are indispensable for its application. For dated references, only the edition cited applies. For undated references, the latest edition of the referenced document (including any amendments) applies.
ISO/IEC Guide 98‑1, Uncertainty of measurement ? Part 1: Introduction to the expression of uncertainty in measurement
ISO/IEC Guide 98‑3, Uncertainty of measurement ? Part 3: Guide to the expression of uncertainty in measurement (GUM:1995)
ISO 20785‑1, Dosimetry for exposures to cosmic radiation in civilian aircraft ? Part 1: Conceptual basis for measurements
ISO 20785‑2, Dosimetry for exposures to cosmic radiation in civilian aircraft ? Part 2: Characterization of instrument response

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ISO 16638-1:2015 specifies the minimum requirements for the design of professional programmes to monitor workers exposed to uranium compounds. It establishes principles for the development of compatible goals and requirements for monitoring programmes and dose assessment for workers occupationally exposed to internal contamination. It establishes procedures and assumptions for risk analysis, monitoring programmes and the standardised interpretation of monitoring data in order to achieve acceptable levels of reliability for uranium and its compounds. It sets limits for the applicability of the procedures in respect to dose levels above which more sophisticated methods have to be applied.
Uranium is both radiologically and chemically toxic. Hence, the scientific bases of current occupational exposure standards are reviewed in addition to radiation exposure limits. This International Standard addresses those circumstances when exposure could be constrained by either radiological or chemical toxicity concerns.
ISO 16638-1:2015 addresses, for uranium and its compounds, the following items:
a)    purposes of monitoring and monitoring programmes;
b)    description of the different categories of monitoring programmes;
c)    quantitative criteria for conducting monitoring programmes;
d)    suitable methods for monitoring and criteria for their selection;
e)    information that has to be collected for the design of a monitoring programme;
f)     general requirements for monitoring programmes (e.g. detection limits, tolerated uncertainties);
g)    frequencies of measurements;
h)    procedures for dose assessment based on reference levels for routine and special monitoring programmes;
i)     assumptions for the selection of dose-critical parameter values;
j)     criteria for determining the significance of monitoring results;
k)    interpretation of workplace monitoring results;
l)     uncertainties arising from dose assessment and interpretation of bioassays data;
m)   reporting/documentation;
n)    quality assurance;
o)    record keeping requirements.
It is not applicable to the following items:
a)    monitoring of exposure due to uranium progeny, including radon;
b)    detailed descriptions of measuring methods and techniques for uranium;
c)    dosimetry for litigation cases;
d)    modelling for the improvement of internal dosimetry;
e)    potential influence of counter-measures (e.g. administration of chelating agents);
f)     investigation of the causes or implications of an exposure;
g)    dosimetry for ingestion exposures and for contaminated wounds.

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ISO 15651:2015 describes a procedure for measuring the total hydrogen content of UO2 and PuO2 powders (up to 2 000 µg/g oxide) and of U02 and (U,Gd)O2 and (U,Pu)O2 pellets (up to 10 µg/g oxide). The total hydrogen content results from adsorbed water, water of crystallization,  hydro-carbon, and other hydrogenated compounds which can exist as impurities in the fuel.

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ISO 21483:2013 specifies an analytical method for determining the solubility in nitric acid of plutonium in pellets of unirradiated mixed oxide fuel (light-water reactor fuels). The results provide information about the expected dissolution behaviour of irradiated pellets under industrial reprocessing conditions. In this aspect, the specific conditions (e.g. time of the test) may vary depending upon the need to match to a specific reprocessor's requirements. The test is aimed at determining solubility under equilibrium conditions rather than the kinetics of dissolution and hence allows for a second dissolution period.

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ISO 15382:2015 provides procedures for monitoring the dose to the skin, the extremities, and the lens of the eye. It gives guidance on how to decide if such dosemeters are needed and to ensure that individual monitoring is appropriate to the nature of the exposure, taking practical considerations into account. National regulations, if they exist, provide requirements that need to be followed.
ISO 15382:2015 specifies procedures for individual monitoring of radiation exposure of the skin, extremities (hands, fingers, wrists, forearms, feet and ankles), and lens of the eye in planned exposure situations. It covers practices which involve a risk of exposure to photons in the range of 8 keV to 10 MeV and electrons and positrons in the range of 60 keV to 10 MeV.
ISO 15382:2015 gives guidance for the design of a monitoring program to ensure compliance with legal individual dose limits. It refers to the appropriate operational dose quantities, and it gives guidance on the type and frequency of individual monitoring and the type and positioning of the dosemeter. Finally, different approaches to assess and analyse skin, extremity, and lens of the eye doses are given.
It is not in the scope of this International Standard to consider exposure due to alpha or neutron radiation fields.

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ISO 20553:2006 specifies the minimum requirements for the design of professional programmes to monitor workers exposed to the risk of internal contamination by radioactive substances and establishes principles for the development of compatible goals and requirements for monitoring programmes.

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ISO 15366-2:2014 describes procedures to chemically separate and purify uranium and plutonium in dissolved solutions of irradiated light water reactor fuels and in samples of high active liquid waste of spent fuel reprocessing plants, prior to their isotopic analysis by e.g. mass spectrometric method or alpha spectrometry. ISO 15366-2:2014 describes a slightly different separation technique from ISO 15366-1, based on the same chemistry, using smaller columns, different support material and special purification steps, applicable to samples containing plutonium and uranium amounts in the nanogram range and below. The detection limits were found to be 500 pg plutonium and 500 pg uranium.

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ISO 15646:2014 describes a procedure for measuring the densification of UO2, (U,Gd)O2, and (U,Pu)O2 pellets, achieved by heat treatment under defined conditions.
The densification of fuel in power operation is an important design feature. Essentially, it is dependent on structural parameters such as pore size, spatial pore distribution, grain size, and in the case of (U,Gd)O2 and (U,Pu)O2, oxide phase structure. A thermal re-sintering test can be used to characterize the dimensional behaviour of the pellets under high temperature. The results of this test are used by the fuel designer to predict dimensional behaviour in the reactor, because thermal densification in the reactor is also dependent on these structural parameters, albeit in a differing manner in terms of quantity.

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ISO 18589-7:2013 specifies the identification of radionuclides and the measurement of their activity in soil using in situ gamma spectrometry with portable systems equipped with germanium or scintillation detectors.
ISO 18589-7:2013 is suitable to rapidly assess the activity of artificial and natural radionuclides deposited on or present in soil layers of large areas of a site under investigation.
ISO 18589-7:2013 can be used in connection with radionuclide measurements of soil samples in the laboratory (ISO 18589‑3) in the following cases:
·        routine surveillance of the impact of radioactivity released from nuclear installations or of the evolution of radioactivity in the region;
·        investigations of accident and incident situations;
·        planning and surveillance of remedial action;
·        decommissioning of installations or the clearance of materials.
It can also be used for the identification of airborne artificial radionuclides, when assessing the exposure levels inside buildings or during waste disposal operations.
Following a nuclear accident, in situ gamma spectrometry is a powerful method for rapid evaluation of the gamma activity deposited onto the soil surface as well as the surficial contamination of flat objects.

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ISO 15366-1:2014 describes procedures to chemically separate and purify uranium and plutonium in dissolved solutions of irradiated light water reactor fuels and in samples of high active liquid waste of spent fuel reprocessing plants, prior to their isotopic analysis by e.g. mass spectrometric method or alpha spectrometry. ISO 15366-1:2014 describes a technique for the separation of uranium and plutonium in spent fuel type samples based on chromatographic method. The procedure applies to samples containing 1 μg to 150 μg Pu (IV) and (VI) and 0,1 mg to 2 mg U (IV) and (VI) in up to 2 ml of 3 mol·l-1 nitric acid solution. It is applicable to mixtures of uranium and plutonium in which the U/Pu-ratio can range from 0 up to 200.

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ISO 16641:2014 covers integrated measurement techniques for radon-220 with passive sampling only. It provides information on measuring the average activity concentration of radon-220 in the air, based on easy-to-use and low-cost passive sampling, and the conditions of use for the measuring devices.
ISO 16641:2014 covers samples taken without interruption over periods varying from a few months to one year.

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Specifies shape, proportions, application, restrictions on the use of the symbol (possibly accompanied by additional symbols or words). Shall be used to signify the actual or potential presence of ionizing radiation (including gamma and X-rays. alpha and beta particles, high-speed electrons, neutrons, protons and other nuclear particles, but not sound waves and other types of electromagnetic waves). Does not specify the radiation levels at which it is to be used.

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ISO 3925:2014 establishes the requirements for the identification and documentation of unsealed radioactive substances issued commercially by suppliers and which are intended for further handling or processing, either physical or chemical. Requirements for radiopharmaceuticals and standard sources are not covered.

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ISO 11665-7:2012 gives guidelines for estimating the radon-222 surface exhalation rate over a short period (a few hours), at a given place, at the interface of the medium (soil, rock, laid building material, walls, etc.) and the atmosphere. This estimation is based on measuring the radon activity concentration emanating from the surface under investigation and accumulated in a container of a known volume for a known duration.
This method is estimative only, as it is difficult to quantify the influence of many parameters in environmental conditions. ISO 11665-7:2012 is particularly applicable, however, in case of an investigation, a search for sources or a comparative study of exhalation rates at the same site. ISO 11665-7:2012 does not cover calibration conditions for the rate estimation devices.

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ISO 2919:2012 establishes a classification system for sealed radioactive sources that is based on test performance and specifies general requirements, performance tests, production tests, marking and certification. It provides a set of tests by which manufacturers of sealed radioactive sources can evaluate the safety of their products in use and users of such sources can select types which are suitable for the required application, especially where protection against the release of radioactive material, with consequent exposure to ionizing radiation, is concerned. ISO 2919:2012 can also serve as guidance to regulating authorities.
The tests fall into several groups, including, for example, exposure to abnormally high and low temperatures and a variety of mechanical tests. Each test can be applied in several degrees of severity. The criterion of pass or fail depends on leakage of the contents of the sealed radioactive source.
Although ISO 2919:2012 classifies sealed sources by a variety of tests, it does not imply that a sealed source will maintain its integrity if used continuously at the rated classification. For example, a sealed source tested for 1 h at 600 °C might, or might not, maintain its integrity if used continuously at 600 °C.
A list of the main typical applications of sealed radioactive sources, with a suggested test schedule for each application, is given in Table 3. The tests constitute minimum requirements corresponding to the applications in the broadest sense. Factors to be considered for applications in especially severe conditions are listed in 4.2.
ISO 2919:2012 makes no attempt to classify the design of sources, their method of construction or their calibration in terms of the radiation emitted. Radioactive materials inside a nuclear reactor, including sealed sources and fuel elements, are not covered by ISO 2919:2012.

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This document describes the principles for the measurement of the activity of 90Sr in equilibrium with 90Y and 89Sr, pure beta emitting radionuclides, in soil samples. Different chemical separation methods are presented to produce strontium and yttrium sources, the activity of which is determined using proportional counters (PC) or liquid scintillation counters (LSC). 90Sr can be obtained from the test samples when the equilibrium between 90Sr and 90Y is reached or through direct 90Y measurement. The selection of the measuring method depends on the origin of the contamination, the characteristics of the soil to be analysed, the required accuracy of measurement and the resources of the available laboratories.
These methods are used for soil monitoring following discharges, whether past or present, accidental or routine, liquid or gaseous. It also covers the monitoring of contamination caused by global nuclear fallout.
In case of recent fallout immediately following a nuclear accident, the contribution of 89Sr to the total amount of strontium activity will not be negligible. This standard provides the measurement method to determine the activity of 90Sr in presence of 89Sr.
The test methods described in this document can also be used to measure the radionuclides in sludge, sediment, construction material and products by following proper sampling procedure.
Using samples sizes of 20 g and counting times of 1 000 min, detection limits of (0,1 to 0,5) Bq·kg-1 can be achievable for 90Sr using conventional and commercially available proportional counter or liquid scintillation counter when the presence of 89Sr can be neglected. If 89Sr is present in the test sample, detection limits of (1 to 2) Bq·kg-1 can be obtained for both 90Sr and 89Sr using the same sample size, counting time and proportional counter or liquid scintillation counter as in the previous situation.

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This document specifies methods and procedures for characterizing the responses of devices used for the determination of ambient dose equivalent for the evaluation of exposure to cosmic radiation in civilian aircraft. The methods and procedures are intended to be understood as minimum requirements.

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This document describes rules for the procedures, applications, and systems of thermoluminescence dosimetry (TLD) for dose measurements according to the probe method. It is particularly applicable to solid "TL detectors", i.e. rods, chips, and microcubes, made from LiF:Mg,Ti or LiF:Mg,Cu,P in crystalline or polycrystalline form. It is not applicable to LiF powders because their use requires special procedures. The probe method encompasses the arrangement, particularly in a water phantom or in a tissue-equivalent phantom, of single TL detectors or of "TL probes", i.e. sets of TL detectors arranged in thin-walled polymethyl methacrylate (PMMA) casings.
The purpose of these rules is to guarantee the reliability and the accuracy indispensable in clinical dosimetry when applied on or in the patient or phantom. This document applies to dosimetry in teletherapy with both photon radiation from 20 keV to 50 MeV and electron radiation from 4 MeV to 25 MeV, as well as in brachytherapy with photon-emitting radionuclides. These applications are complementary to the use of ionization chambers.

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The use of a continuous air monitor (CAM) is mainly motivated by the need to be alerted quickly and in the most accurate way possible with an acceptable false alarm rate when a significant activity concentration value is exceeded, in order to take appropriate measures to reduce exposure of those involved.
The performance of this CAM does not only depend on the metrological aspect characterized by the decision threshold, the limit of detection and the measurement uncertainties but also on its dynamic capacity characterized by its response time as well as on the minimum detectable activity concentration corresponding to an acceptable false alarm rate.
The ideal performance is to have a minimum detectable activity concentration as low as possible associated with a very short response time, but unfortunately these two criteria are in opposition. It is therefore important that the CAM and the choice of the adjustment parameters and the alarm levels be in line with the radiation protection objectives.
The knowledge of a few factors is needed to interpret the response of a CAM and to select the appropriate CAM type and its operating parameters.
Among those factors, it is important to know the half-lives of the radionuclides involved, in order to select the appropriate detection system and its associated model of evaluation.
CAM using filter media accumulation sampling techniques are usually of two types:
a)    fixed filter;
b)    moving filter.
This document first describes the theory of operation of each CAM type i.e.:
—     the different models of evaluation considering short or long radionuclides half-lives values,
—     the dynamic behaviour and the determination of the response time.
In most case, CAM is used when radionuclides with important radiotoxicities are involved (small value of ALI). Those radionuclides have usually long half-life values.
Then the determination of the characteristic limits (decision threshold, detection limit, limits of the coverage interval) of a CAM is described by the use of long half-life models of evaluation.
Finally, a possible way to determine the minimum detectable activity concentration and the alarms setup is pointed out.
The annexes of this document show actual examples of CAM data which illustrate how to quantify the CAM performance by determining the response time, the characteristics limits, the minimum detectable activity concentration and the alarms setup.

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This document specifies additional procedures and data for the calibration of dosemeters and doserate meters used for individual and area monitoring in radiation protection. The general procedure for the calibration and the determination of the response of radiation protection dose(rate)meters is described in ISO 29661 and is followed as far as possible. For this purpose, the photon reference radiation fields with mean energies between 8 keV and 9 MeV, as specified in ISO 4037-1, are used. In Annex D some additional information on reference conditions, required standard test conditions and effects associated with electron ranges are given. For individual monitoring, both whole body and extremity dosemeters are covered and for area monitoring, both portable and installed dose(rate)meters are covered.
Charged particle equilibrium is needed for the reference fields although this is not always established in the workplace fields for which the dosemeter should be calibrated. This is especially true at photon energies without inherent charged particle equilibrium at the reference depth d, which depends on the actual combination of energy and reference depth d. Electrons of energies above 65 keV, 0,75 MeV and 2,1 MeV can just penetrate 0,07 mm, 3 mm and 10 mm of ICRU tissue, respectively, and the radiation qualities with photon energies above these values are considered as radiation qualities without inherent charged particle equilibrium for the quantities defined at these depths. This document also deals with the determination of the response as a function of photon energy and angle of radiation incidence. Such measurements can represent part of a type test in the course of which the effect of further influence quantities on the response is examined.
This document is only applicable for air kerma rates above 1 µGy/h.
This document does not cover the in-situ calibration of fixed installed area dosemeters.
The procedures to be followed for the different types of dosemeters are described. Recommendations are given on the phantom to be used and on the conversion coefficients to be applied. Recommended conversion coefficients are only given for matched reference radiation fields, which are specified in ISO 4037-1:2019, Clauses 4 to 6. ISO 4037‑1:2019, Annexes A and B, both informative, include fluorescent radiations, the gamma radiation of the radionuclide 241Am, S-Am, for which detailed published information is not available. ISO 4037-1:2019, Annex C, gives additional X radiation fields, which are specified by the quality index. For all these radiation qualities, conversion coefficients are given in Annexes A to C, but only as a rough estimate as the overall uncertainty of these conversion coefficients in practical reference radiation fields is not known.
NOTE       The term dosemeter is used as a generic term denoting any dose or doserate meter for individual or area monitoring.

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This document gives guidelines on additional aspects of the characterization of low energy photon radiations and on the procedures for calibration and determination of the response of area and personal dose(rate)meters as a function of photon energy and angle of incidence. This document concentrates on the accurate determination of conversion coefficients from air kerma to Hp(10), H*(10), Hp(3) and H'(3) and for the spectra of low energy photon radiations. As an alternative to the use of conversion coefficients the direct calibration in terms of these quantities by means of appropriate reference instruments is described.

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This document specifies a method of determining the apparent density and tap density of free-flowing uranium dioxide (UO2) powder which will be used for pelleting and sintering of UO2 pellets as a nuclear fuel.
This method can be used for different UO2 powder types including grains, granules, spheres or other kinds of particles. The method can also be applied to other fuel powders as PuO2, ThO2 and powder mixtures as UO2-PuO2 and UO2-Gd2O3.
This document is based on the principle of using a flowmeter funnel (see 4.1). Other measurement apparatus, such as a Scott volumeter, can also be used.

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