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Browsing ANS - American Nuclear Society by Title
Now showing items 1-20 of 87
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ANS 1
Abstract: This standard provides for the safe conduct of critical experiments. Such experiments study neutron behavior in a fission device where the energy produced is insufficient to require auxiliary cooling, and the power history ... -
ANS 10.2
Abstract: This standard provides recommended programming practices and requirements to facilitate the portability of computer programs prefared for scientific and engineering computations. -
ANS 10.4
Abstract: This standard provides guidelines for the verification and validation ~V&V! of non-safetyrelated scientific and engineering computer programs developed for use by the nuclear industry. The scope is restricted to research ... -
ANS 10.5
Abstract: This standard presents criteria for accommodating user needs in the preparation of computer software for scientific and engineering applications. -
ANS 10.7
Abstract: This standard provides minimum requirements for assurance that high-integrity software developed for use by the nuclear industry meets state-of-the-practice expectations for quality. The requirements in this standard are ... -
ANS 14.1
Abstract: This standard is for those involved in the design, operation, and review of fast pulse reactors. It has been formulated in general terms to be applicable to all current fast pulse reactors. This standard does not apply to ... -
ANS 15.1
Abstract: This standard identifies and establishes the content of technical specifications (TS) for research and test reactors. Areas addressed are definitions, safety limits, limiting safety system settings, limiting conditions for ... -
ANS 15.11
Abstract: This standard establishes the elements of a radiation protection program and the criteria necessary to provide an acceptable level of radiation protection for personnel at research reactor facilities and the public consistent ... -
ANS 15.11
Abstract: Scope: This standard establishes the elements of a radiation protection program and the criteria necessary to provide an acceptable level of radiation protection for personnel at research reactor facilities and the public ... -
ANS 15.16
Abstract: This standard identifies the elements of an emergency plan that describes the approach to coping with emergencies and minimizing the consequences of accidents at research reactor facilities. The emphasis given each of these ... -
ANS 15.21
Abstract: This standard provides the criteria for the format and content for safety analysis reports ~SARs! for research reactors.
Purpose
Research reactors require safety analysis and documentation showing ... -
ANS 15.4
Abstract: This standard provides criteria for the selection and training of research reactor operating personnel. It addresses their qualifications, training, initial licensing, requalification, and relicensing.
This standard ... -
ANS 15.4
Abstract: Scope: This standard provides criteria for the selection and training of research reactor operating personnel. It addresses their qualifications, training, initial licensing, requalification, and relicensing. This standard ... -
ANS 15.8
Abstract: This standard provides criteria for quality assurance in the design, construction, operation, and decommissioning of research reactors. -
ANS 16.1
Abstract: This standard ANSI/ANS-16.1-2003,1) provides a unform procedure to measure and index the release of radionuclides from waste forms as a result of leaching in demineralised water for 5 days.2) The ... -
ANS 18.1
Abstract: Scope: This standard provides a set of typical radionuclide concentrations for estimating the radioactivity in the principal fluid systems of light water reactors and for projecting the expected releases of radioactivity ... -
ANS 19.1
Abstract: General
This standard identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include ~a! ... -
ANS 19.10
Abstract: This standard provides a procedure for the evaluation of the best-estimate fast (E > 1.0 MeV) neutron fluence in the annular region between the core and the inside surface of the vessel, through the pressure vessel ... -
ANS 19.11
Abstract: This standard provides guidance and specifies criteria for determining the MTC in water moderated power reactors. Measurement of the isothermal temperature coefficient of reactivity (ITC) at hot zero power (HZP) conditions ...