Mathematics and Computation (previously ANS-10)

  • ANS-10.2, Portability of Scientific and Engineering Software
    Scope: This standard provides recommended programming practices and requirements to facilitate the portability of computer programs prepared for scientific and engineering computations.
  • ANS-10.4, Verification and Validation of Non-Safety-Related Scientific and Engineering Computer Programs for the Nuclear Industry
    Scope: This standard provides guidelines for the verification and validation (V&V) of non-safety-related scientific and engineering computer programs developed for use by the nuclear industry. The scope is restricted to research and other non-safety-related, noncritical applications.
  • ANS-10.5, Accommodating User Needs in Scientific and Engineering Computer Software Development
    Scope: This standard presents criteria for accommodating user needs in the preparation of computer software for scientific and engineering applications.
  • ANS-10.7, Non-Real-Time, High-Integrity Software for the Nuclear Industry--Developer Requirements
    Scope: 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 specified for development of high-integrity software. The intent is to set a minimum level of quality assurance and critical technical process requirements to satisfy due diligence. NUREG/CR 6263 [1],1) from which many of the requirements of this standard are derived, was developed for application to nuclear power plants, and this standard is primarily applicable to nuclear power plants and other nuclear facilities and operations with similar high consequences and hazards. This standard addresses rigorous, systematic development of high-integrity, non-real-time safety analysis, design, and simulation software that includes calculations or simulations requiring high functional reliability in order to avoid undetected errors that could have serious consequences if such errors are not detected (the scope of this standard does not include electronic procedures). It is especially important that this standard be followed in cases where calculations are so complex that typical peer reviews are not likely to identify errors. For complex software, hand calculations and code-to-code comparisons may not be adequate to verify and validate the software. This may include software used for nuclear design and analysis; analysis of postulated accidents and assignment of safety classification levels to systems, structures, and components at nuclear facilities; computational fluid dynamics (CFD); thermal hydraulics; structural mechanics; complex Monte Carlo simulations; radiation dosimetry; and nuclear medical applications. An important area covered in this standard is model development and verification (including physics validation), which are critically important tasks for high-integrity analysis and simulation software. The requirements in this standard for model development and verification take into consideration several standards for the development of computational models and methods, including ANSI/ASME V&V 10-2006 [2]; AIAA G-077-1998 (2002) [3]; U.S. Nuclear Regulatory Commission (NRC) Standard Review Plan, NUREG-0800, Sec. 15.0.2 [4]; and NRC Regulatory Guide 1.203 [5]. This standard provides the requirements necessary to validate the model by specifying requirements for model development and validation, except that it does not address the actual planning, design, and conduct of validation tests/experiments. Cybersecurity is another important aspect of high-integrity software and is explicitly addressed in this standard. The requirements in this standard for security requirements were principally derived from NRC Regulatory Guide 1.152 [6].
  • ANS-10.8, Non-Real Time, High-Integrity Software for the Nuclear Industry--User Requirements
    Scope: This standard establishes the minimum requirements for the acceptance and use of non-real-time, high-integrity software used for design and analysis in the nuclear industry. This standard is directly related to ANSI/ANS-10.7-2013 [1], ) which provides requirements for the developer of non-real-time, high-integrity software. The activities described in this standard enable an end user of complex simulation software to assure, when software has been developed in accordance with appropriate requirements [1], that the software will meet the specific identified needs of the end user and that the software is installed and used in a correct manner. The type of software discussed in this standard is software used for the analysis, design, and simulation of complex physical systems and phenomena. This type of software requires a high degree of functional reliability in order to reduce the risk of undetected errors.

Reactor Physics (previously ANS-19)

  • ANS-5.1, Decay Heat Power in Light Water Reactors
    Scope: This standard sets forth values for the decay heat power from fission products and 239U and 239Np following shutdown of light water reactors (LWRs) containing 235U, 238U, and plutonium. The decay heat power from fission products is presented in tables and equivalent analytical representations. Methods are described that account for the reactor operating history, for the effect of neutron capture in fission products, and for assessing the uncertainty in the resultant decay heat power. Decay heat power from other actinides and activation products in structural materials, and fission power from delayed neutron-induced fission, are not included in this standard and shall be evaluated by the user and appropriately included in any analysis of shutdown power.
  • ANS-19.1, Nuclear Data Sets for Reactor Design Calculations
    Scope: 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) criteria for acceptance of evaluated nuclear data sets, (b) criteria for processing evaluated data and preparation of processed continuous data and averaged data sets (c) identification of specific evaluated, processed continuous and averaged data sets that meet these criteria for specific reactor types.
  • ANS-19.3, Determination of Steady-State Neutron Reaction-Rate Distributions and Reactivity of Nuclear Power Reactors
    Scope: This standard provides guidance for performing and validating the sequence of steady state calculations leading to prediction, in all types of nuclear reactors, of: (1) Reaction rate spatial distributions (2) Reactivity (3) Change of isotopic compositions with time. The standard provides: (1) Guidance for the selection of computational methods. (2) Criteria for verification of calculational methods used by reactor core analysts (3) Criteria for evaluation of accuracy and range of applicability of data and methods (4) Requirements for documentation of the preceding. The scope of the standard is shown schematically in Figure 1.
  • ANS-19.3.4, The Determination of Thermal Energy Deposition Rates in Nuclear Reactors
    Scope: It is the purpose of this standard to provide criteria for: (1) Determination of the energy allocation among the principal particles and photons produced in fission, both prompt and delayed; (2) Adoption of appropriate treatment of heavy charged particle and electron slowing down in matter; (3) Determination of the spatial energy deposition rates resulting from the interactions of neutrons; (4) Calculation of the spatial energy deposition rates resulting from the various interactions of photons with matter; and (5) Presentation of the results of such computations, including verification of accuracy and specification of uncertainty. This standard addresses the energy generation and deposition rates for all types of nuclear reactors where the neutron reaction rate distribution and photon and beta emitter distributions are known. Its scope is limited to the reactor core, including blanket zones, control elements and core internals, pressure vessel, and the thermal and biological shielding.
  • ANS-19.4, A Guide for Acquisition and Documentation of Reference Power Reactor Physics Measurements for Nuclear Analysis Verification
    Scope: This standard applies to measurements of reactor parameters in light water power reactors that are intended to serve as reference measurements to be used in evaluating reactor physics computational procedures. It includes: identification of the types of parameters of interest as reference measurements; a brief description of test conditions and experimental data required for such reference measurements; identification of problems and concerns which may affect the accuracy or interpretation of the data; and criteria to be used in documenting the results of reference measurements.
  • ANS-19.5, Requirements for Reference Reactor Physics Measurements
    Scope: This standard provides criteria for the qualification of reference reactor physics measurements obtained from subcritical (including non-multiplying), critical, and other experiments for the purpose of verifying nuclear design and analysis methods. It also provides criteria for documentation of reference data and review of proposed reference reactor physics data to ensure compliance with this standard.
  • ANS-19.6.1, Reload Startup Physics Tests for Pressurized Water Reactors
    Scope: This standard specifies the minimum acceptable startup reactor physics test program to determine if the operating characteristics of the core are consistent with the design predictions, which provides assurance that the core can be operated as designed.
  • ANS-19.8, Fission Product Yields for 235U, 238U, and 239Pu
    Scope: This standard provides a reference set of fission yield data for thermal and fast neutron-induced fission of 233 U, 235U, 239Pu, and 241Pu; fast neutron-fission of 232Th, 238U, and 240Pu; and spontaneous fission of 252Cf. The data for these 12 fissioning systems are given as mass chain yields and their uncertainties and are presented in tabular form. Discussions are presented and references given concerning the application of the data. Concerns associated with the uncertainties in the mass chain yields are also discussed. A set of cumulative fission yields and uncertainties are included explicitly for a number of special purpose fission-product nuclides, particularly those important to dosimetry.
  • ANS-19.9, Delayed Neutron Parameters for Light Water Reactors
    Scope: This standard provides energy-dependent delayed neutron yield and decay data for Light Water Reactor design and control. The standard addresses the identification and characterization of fission products leading to delayed neutron emission; the total delayed neutron yield as a function of energy for U-233, U-235, U-238 and Pu-239; and fractions associated with individual emitters, half-lives and spectra for the classical group representation of delayed neutron data
  • ANS-19.10, Methods for Determining Neutron Fluence in BWR and PWR Pressure Vessel and Reactor Internals
    Scope: This standard provides criteria for performing and validating the sequence of calculations required for the prediction of the fast neutron fluence t in the reactor vessel. Applicable to PWR and BWR plants the standard addresses flux attenuation from the core through the vessel to the cavity and provides criteria for generating cross sections, spectra, transport and comparisons with in- and ex-vessel measurements, validation, uncertainties and flux extrapolation to the inside vessel surface.
  • ANS-19.11, Calculation and Measurement of the Moderator Temperature Coefficient of Reactivity for Pressurized Water Reactors
    Scope: This standard provides guidance and specifies criteria for determining the MTC in PWRs. Measurement of the isothermal temperature coefficient of reactivity (ITC) at hot-zero-power (HZP) conditions is covered in ANSI/ANS-19.6.1-2011 (R2016). This standard therefore addresses the calculation of the ITC at HZP and the calculation and measurement of the MTC at power. This standard addresses the calculation and measurement of the MTC only in PWRs because that is the only type of power reactor currently sited in the United States for which measurement of the MTC is required.
  • ANS-19.12, Nuclear Data for Isotope Production Calculations for Medical and Other Applications
    Scope: This standard establishes criteria for developing evaluated neutron cross section and branching ratio data for isotope production pathways for fast and thermal reactor systems, providing the data needed to calculate production of the desired medical and other isotopes and associated impurities.
  • ANS-19.13, Initial Fuel Loading and Startup Tests for FOAK Advanced Reactors
    Scope: This standard provides best practices for reactor startup of First-of-a-Kind (FOAK) Advanced Reactors (AR) such that basic safety, operational, and fundamental property data for technical and safety specifications confirmation, and design software validation, can be performed concurrently. Best practices for startup of heritage reactors and modern light water reactors will be assimilated into generic recommended startup procedures for future FOAK-ARs. This standard will provide traceability between such recommended best practices and the identified key integral data. It will thus allow auditing the methodology of new FOAK ARs.

Shielding (previously ANS-6)

  • ANS-5.4, Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel
    Scope: This standard provides an analytical method for calculating the release of volatile fission products from oxide fuel pellets during normal reactor operation. When used with nuclide yields, this method will give the so-called "gap activity," which is the inventory of volatile fission products that could be available for release from the fuel rod if the cladding were breached. The standard considers high-temperature (up to the melting point) and low-temperature (where temperature-independent processes dominate) releases and distinguishes between short-halflife (halflife less than one year) and long-halflife (halflife greater than one year) nuclides. This standard requires that releases for nuclides of interest be calculated with both the high-temperature and the low-temperature models, and the larger of the two calculated releases is to be taken as the result.
  • ANS-5.10, Airborne Release Fractions at Non-Reactor Nuclear Facilities
    Scope: This standard provides criteria for defining Airborne Release Fractions (ARFs) for radioactive materials under accident conditions (excluding nuclear criticalities) at non-reactor nuclear facilities. The criteria in this standard provide requirements for selecting ARFs based on the calculated or assumed forms of radioactive material released. This standard may be applied to determine the ARFs for certain applicable reactor plant events for which alternative methodologies are not mandated by regulatory requirements. Because the predominant physical forms of radioactive materials in non-reactor facilities are solids and liquids, the standard focuses on these forms. Criteria are also provided for gases and materials that can be converted into the form of a vapor.
  • ANS-6.1.1, Neutron and Photon Fluence-to-Dose Conversion Coefficients
    Scope: This standard presents data recommended for computing the biologically relevant dosimetric quantity in photon and neutron radiation fields. Specifically, this standard is intended for use by radiation shielding designers for the calculation of effective dose. Fit coefficients are given for evaluating whole body effective dose per unit fluence for photons with energy between 10 keV to 10 GeV and for neutrons with energy between 0.001 eV to 10 GeV. Eight different irradiation geometries are considered. Establishing exposure limits is outside the scope of this standard.
  • ANS-6.1.2, Group-Averaged Neutron and Gamma-Ray Cross Sections for Radiation Protection and Shielding Calculations for Nuclear Power Plants
    Scope: This standard provides information on acceptable evaluated nuclear data and group-averaged neutron and gamma-ray cross section libraries derived from these evaluated nuclear data based on the energy range and materials of importance in nuclear radiation protection and shielding calculations for nuclear power plants.
  • ANS-6.3.1, Program for Testing Radiation Shields in Light Water Reactors (LWR)
    Scope: This standard describes a test program to be used in evaluating biological radiation shielding in nuclear reactor facilities under normal operating conditions including anticipated operational occurrences. The program encompasses examining and testing to be performed before startup, during startup, and testing subsequent to the startup phase. Post startup tests are required for the shielded components which do not contain sufficient radioactivity during the startup phase to allow valid testing. Shielding of these components is to be tested when radiation sources develop or are introduced into sufficient strength to allow meaningful measurements. Post startup shield tests are also required whenever radioactive or potentially radioactive equipment which could affect the adequacy of the installed shielding is introduced into the plant or relocated within the plant, or when previously tested shielding has been modified. One special category of post start-up testing is the testing of shielding during refueling operations.
  • ANS-6.4, Nuclear Analysis and Design of Concrete Radiation Shielding for Nuclear Power Plants
    Scope: The standard contains methods and data needed in design of concrete shielding required for protection of personnel and equipment against the effects of gamma rays and neutrons. Specific guidance is given regarding attenuation calculations, shielding design, and standards of documentation.
  • ANS-6.4.2, Specification for Radiation Shielding Materials
    Scope: This standard sets forth physical and nuclear properties that shall be reported by the supplier as appropriate for a particular application in order to form the basis for the selection of radiation shielding materials.
  • ANS-6.4.3, Gamma-Ray Attenuation Coefficients & Buildup Factors for Engineering Materials
    Scope: This standard presents evaluated gamma-ray elemental attenuation coefficients and single-material buildup factors for selected engineering materials for use in shielding calculations of structures in power plants and other nuclear facilities. The data cover the energy range 0.015-15 MeV and up to 40 mean free paths (mfp). These data are intended to be standard reference data for use in radiation analyses employing point-kernel methods.
  • ANS-6.6.1, Calculation and Measurement of Direct and Scattered Gamma Radiation from LWR Nuclear Power Plants
    Scope: This standard defines calculational requirements and discusses measurement techniques for estimates of dose rates near light water reactor (LWR) nuclear power plants due to direct and scattered gamma-rays from contained sources onsite. Onsite locations outside plant buildings and locations in the offsite unrestricted area are considered. All sources that contribute significantly to dose rates are identified and methods for calculating the source strength of each are discussed. Particular emphasis is placed on 16N sources as they are significant sources of direct and scattered radiation for boiling water reactors (BWR). The standard specifically excludes radiation from gaseous and liquid effluents. The standard describes the considerations necessary to compute dose rates, including component self-shielding, shielding afforded by walls and structures, and scattered radiation. The requirements for measurements and data interpretation of measurements are given. The standard includes normal operation and shutdown conditions but does not address accident or normal operational transient conditions.