1.1 This specification covers uranium oxides, including processed byproducts or scrap material (powder, pellets, or pieces), that are intended for dissolution into uranyl nitrate solution meeting the requirements of Specification C788 prior to conversion into nuclear grade UO2 powder with a 235U content of less than 5 %. This specification defines the impurity and uranium isotope limits for such urania powders that are to be dissolved prior to processing to nuclear grade UO2 as defined in Specification C753.1.2 This specification provides the nuclear industry with a general standard for such uranium oxide powders. It recognizes the diversity of conversion processes and the processes to which such powders are subsequently to be subjected (for instance, by solvent extraction). It is therefore anticipated that it may be necessary to include supplementary specification limits by agreement between the buyer and seller.1.3 The scope of this specification does not comprehensively cover all provisions for preventing criticality accidents, for health and safety, or for shipping. Observance of this specification does not relieve the user of the obligation to conform to all international, national, state and local regulations for processing, shipping, or any other way of using urania powders (see 2.2 and 2.3).1.4 Units—The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.1.5 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.

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4.1 This standard guide applies to concrete that is still in place with a defined geometry and known, documented history.4.2 It is not intended for use on concrete that has already been rubbelized where it is difficult to measure the radiation levels and not easy to remove surface contamination to reduce radiation levels after concrete has been rubbelized.4.3 This standard guide applies to surface or volumetrically contaminated concrete, where the depth of contamination can be measured or estimated based on the history of the concrete.4.4 This standard guide does not apply to the reinforcement bar (rebar) found in concrete. Although most concrete contains rebar, it is generally removed before the concrete is dispositioned. In addition, rebar may be activated, and is covered under procedures for reuse of scrap metal.4.5 General unit-dose and unit-cost data to support the calculations is provided in the appendices of this standard guide. However, if site-specific data is available, it should be used instead of the general information provided here.4.6 This standard guide helps determine estimated doses to the public during disposal of concrete and to future residents of disposal areas. It does not include dose to radiation workers already involved in a radiation control program. It is assumed that the dose to radiation workers is already tracked and kept within acceptable levels through a radiation control program. The cost and dose to radiation workers could be added in to find an overall cost and dose for each option.1.1 This standard guide defines the process for developing a strategy for dispositioning concrete from nuclear facility decommissioning. It outlines a 10-step method to evaluate disposal options for radioactively contaminated concrete. One of the steps is to complete a detailed analysis of the cost and dose to nonradiation workers (the public); the methodology and supporting data to perform this analysis are detailed in the appendices. The resulting data can be used to balance dose and cost and select the best disposal option. These data, which establish a technical basis to apply to release the concrete, can be used in several ways: (1) to show that the release meets existing release criteria, (2) to establish a basis to request release of the concrete on a case-by-case basis, (3) to develop a basis for establishing release criteria where none exists.1.2 This standard guide is based on the “Protocol for Development of Authorized Release Limits for Concrete at U.S. Department of Energy Sites,” (1)2 from which the analysis methodology and supporting data are taken.1.3 Guide E1760 provides a general process for release of materials containing residual amounts of radioactivity. In addition, Guide E1278 provides a general process for analyzing radioactive pathways. This standard guide is intended for use in conjunction with Guides E1760 and E1278, and provides a more detailed approach for the release of concrete.1.4 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.

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This test method uses a high-resolution gamma-ray spectrometer as a basis for measuring the gamma-ray emission rate of 137Cs-137mBa in a dilute nitric acid solution containing 10 mg/L of cesium carrier. No chemical separation of the cesium from the dissolved-fuel solution is required. The principal steps consist of diluting a weighed aliquot of the dissolved-fuel solution with a known mass of 1 M nitric acid (HNO3) and measuring the 662 keV gamma-ray count rate from the sample, then measuring the 662 keV gamma-ray count rate from a standard source that has the same physical form and counting geometry as the sample.The amount of fuel sample required for the analysis is small. For a sample containing 0.1 g of fuel irradiated to one atom percent fission, a net count rate of approximately 105 counts per second will be observed for a counting geometry that yields a full-energy peak efficiency fraction of 1 × 10-3. The advantage of this small amount of sample is that the concentration of fuel material can be kept at levels well below 1 g/L, which results in negligible self-absorption in the sample aliquot and a small radiation hazard to the analyst.1.1 This test method covers the determination of the number of atoms of 137Cs in aqueous solutions of irradiated uranium and plutonium nuclear fuel. When combined with a method for determining the initial number of fissile atoms in the fuel, the results of this analysis allows atom percent fission (burn-up) to be calculated (1). The determination of atom percent fission, uranium and plutonium concentrations, and isotopic abundances are covered in Test Methods E 267 and E 321.1.2 137Cs is not suitable as a fission monitor for samples that may have lost cesium during reactor operation. For example, a large temperature gradient enhances 137Cs migration from the fuel region to cooler regions such as the radial fuel-clad gap, or, to a lesser extent, towards the axial fuel end.1.3 A nonuniform 137Cs distribution should alert the analyst to the potential loss of the fission product nuclide. The 137Cs distribution may be ascertained by an axial gamma-ray scan of the fuel element to be assayed. In a mixed-oxide fuel, comparison of the 137Cs distribution with the distribution of nonmigrating fission-product nuclides such as 95Zr or 144Ce would indicate the relative degree of 137Cs migration.1.4 The values stated in SI units are to be regarded as standard. No other unites of measurement are included in this standard.1.5 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.

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1.1 This guide is intended to assist the maintenance engineer in the preparation of a specification or work instruction for re-coating items that are presently coated with what is known within the nuclear power industry as an "unqualified coating."

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4.1 Establishment of an in-service coatings monitoring program permits planning and prioritization of coatings maintenance work as needed to maintain coating integrity and performance in nuclear CSL I coating systems. For additional information on nuclear maintenance coating work, refer to ASTM MNL8.44.2 A coatings monitoring program enables early identification and detection of potential problems in coating systems. Some CSL I coating systems may be known in advance to be suspect, deficient, or unqualified. Monitoring coating performance will assist in developing follow-up procedures to resolve any significant deficiency relative to coating work.4.3 Degraded coatings may generate debris under design basis accident conditions that could adversely affect the performance of the post-accident safety systems. A coatings monitoring program may be required to fulfill safety analysis report and generic letter commitments for CSL I coating work in a nuclear power plant facility.1.1 This standard covers procedures for establishing a monitoring program for condition assessment of Coating Service Level (CSL) I coating systems in operating nuclear power plants. Monitoring is an ongoing process of evaluating the condition and performance of the in-service coating systems.1.2 It is the intent of this standard to provide a recommended basis for establishing a coatings condition assessment program, not to mandate a singular basis for all programs. Variations or simplifications of the program described in this standard may be appropriate for each operating nuclear power plant depending on their licensing commitments.1.3 This requirements of ASME Section XI, In-Service Inspection Subsections IWE and IWL are beyond the scope of 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.1.5 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.

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