ASTM E482-16

3.1 General: 

3.2 Validation—Prior to performing transport calculations for a particular facility, the computational methods must be validated by comparing results with measurements made on a benchmark experiment. Criteria for establishing a benchmark experiment for the purpose of validating neutronics methodology should include those set forth in Guides E944 and E2006 as well as those prescribed in 3.2.1. A discussion of the limiting accuracy of benchmark validation discrete ordinate radiation transport procedures for the LWR surveillance program is given in Ref 3.2.1 Requirements for Benchmarks—In order for a particular experiment to qualify as a calculational benchmark, the following criteria are recommended:

3.2.2 Methodology Validation—It is essential that the neutronics methodology employed for predicting neutron fluence in a reactor pressure vessel be validated by accurately predicting appropriate benchmark dosimetry results. In addition, the following documentation should be submitted: (1) convergence study results, and (3.2.2.1 For example, model specifications for discrete-ordinates method on which convergence studies should be performed include: (1) neutron cross-sections or energy group structure, (3.2.2.2 Uncertainties that are propagated from known uncertainties in nuclear data need to be addressed in the analysis. The uncertainty analysis for discrete ordinates codes may be performed with sensitivity analysis as discussed in References 3.2.2.3 Effects of known uncertainties in geometry and source distribution should be evaluated based on the following test cases: (1) reference calculation with a time-averaged source distribution and with best estimates of the core, and pressure vessel locations, (2) reference case geometry with maximum and minimum expected deviations in the source distribution, and (3.2.2.4 Measured and calculated integral parameters should be compared for all test cases. It is expected that larger uncertainties are associated with geometry and neutron source specifications than with parameters included in the convergence study. Problems associated with space, energy, and angle discretizations can be identified and corrected. Uncertainties associated with geometry specifications are inherent in the structure tolerances. Calculations based on the expected extremes provide a measure of the sensitivity of integral parameters to the selected variables. Variations in the proposed convergence and uncertainty evaluations are appropriate when the above procedures are inconsistent with the methodology to be validated. As-built data could be used to reduce the uncertainty in geometrical dimensions.

3.2.2.5 In order to illustrate quantitative criteria based on measurements and calculations that should be satisfied, let ψ denote a set of logarithms of calculation (where q_i and and the best estimate of the variance, 3.2.2.6 The neutronics methodology is validated, if (in addition to qualitative model evaluation) all of the following criteria are satisfied:

3.2.2.7 Note that a nonzero log-mean of the 3.2.2.8 One acceptable procedure for performing these comparisons is: (1) obtain group fluence rates at dosimeter locations from neutronics calculations, (2) collapse the Guide E1018 recommended dosimetry cross section data to a multigroup set consistent with the neutron energy group fluence rates or obtain a fine group spectrum (consistent with the dosimetry cross section data) from the calculated group fluence rates, (3.3 Determination of the Fixed Fission Source—The power distribution in a typical power reactor undergoes significant change during the life of the reactor. A time-averaged power distribution is recommended for use in determination of the neutron source distribution utilized for damage predictions. An adjoint procedure, described in 3.3.2, may be more appropriate for dosimetry comparisons involving product nuclides with short half-lives. For multigroup methods, the fixed source may be determined from the equation: