WSRC Approach to Validation of Criticality Safety Computer Codes

WSRC Approach to Validation of Criticality Safety Computer Codes

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Published: 1991

Total Pages: 11

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Recent hardware and operating system changes at Westinghouse Savannah River Site (WSRC) have necessitated review of the validation for JOSHUA criticality safety computer codes. As part of the planning for this effort, a policy for validation of JOSHUA and other criticality safety codes has been developed. This policy will be illustrated with the steps being taken at WSRC. The objective in validating a specific computational method is to reliably correlate its calculated neutron multiplication factor (K{sub eff}) with known values over a well-defined set of neutronic conditions. Said another way, such correlations should be: (1) repeatable; (2) demonstrated with defined confidence; and (3) identify the range of neutronic conditions (area of applicability) for which the correlations are valid. The general approach to validation of computational methods at WSRC must encompass a large number of diverse types of fissile material processes in different operations. Special problems are presented in validating computational methods when very few experiments are available (such as for enriched uranium systems with principal second isotope 236U). To cover all process conditions at WSRC, a broad validation approach has been used. Broad validation is based upon calculation of many experiments to span all possible ranges of reflection, nuclide concentrations, moderation ratios, etc. Narrow validation, in comparison, relies on calculations of a few experiments very near anticipated worst-case process conditions. The methods and problems of broad validation are discussed.


Certification Process of Safety Analysis and Risk Management Computer Codes at the Savannah River Site

Certification Process of Safety Analysis and Risk Management Computer Codes at the Savannah River Site

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Published: 1992

Total Pages: 10

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The commitment by Westinghouse Savannah River Company (WSRC) to bring safety analysis and risk management codes into compliance with national and sitewide quality assurance requirements necessitated a systematic, structured approach. As a part of this effort, WSRC, in cooperation with the Westinghouse Hanford Company, has developed and implemented a certification process for the development and control of computer software. Safety analysis and risk management computer codes pertinent to reactor analyses were selected for inclusion in the certification process. As a first step, documented plans were developed for implementing verification and validation of the codes, and establishing configuration control. User qualification guidelines were determined. The plans were followed with an extensive assessment of the codes with respect to certification status. Detailed schedules and work plans were thus determined for completing certification of the codes considered. Although the software certification process discussed is specific to the application described, it is sufficiently general to provide useful insights and guidance for certification of other software.


Criticality Safety Validation: Simple Geometry, Single Unit[sup 233]U Systems

Criticality Safety Validation: Simple Geometry, Single Unit[sup 233]U Systems

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Published: 2001

Total Pages:

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Typically used LMITCO criticality safety computational methods are evaluated for suitability when applied to INEEL[sup 233]U systems which reasonably can be modeled as simple-geometry, single-unit systems. Sixty-seven critical experiments of uranium highly enriched in[sup 233]U, including 57 aqueous solution, thermal-energy systems and 10 metal, fast-energy systems, were modeled. These experiments include 41 cylindrical and 26 spherical cores, and 41 reflected and 26 unreflected systems. No experiments were found for intermediate-neutron-energy ranges, or with interstitial non-hydrogenous materials typical of waste systems, mixed[sup 233]U and plutonium, or reflectors such as steel, lead, or concrete. No simple geometry experiments were found with cubic or annular cores, or approximating infinite sea systems. Calculations were performed with various tools and methodologies. Nine cross-section libraries, based on ENDF/B-IV, -V, or -VI. 2, or on Hansen-Roach source data, were used with cross-section processing methods of MCNP or SCALE. The k[sub eff] calculations were performed with neutral-particle transport and Monte Carlo methods of criticality codes DANT, MCNP 4A, and KENO Va.


Validation and Verification Plan for Safety and PRA Codes

Validation and Verification Plan for Safety and PRA Codes

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Published: 1991

Total Pages: 47

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This report discusses a verification and validation (V V) plan for computer codes used for safety analysis and probabilistic risk assessment calculations. The present plan fulfills the commitments by Westinghouse Savannah River Company (WSRC) to the Department of Energy Savannah River Office (DOE-SRO) to bring the essential safety analysis and probabilistic risk assessment codes in compliance with verification and validation requirements.


Bias and Uncertainty of Critical Experiment Models with CSAS25 from SCALE4.4a for Criticality Safety Analyses On the HP J-5600 (CMODB) Workstation

Bias and Uncertainty of Critical Experiment Models with CSAS25 from SCALE4.4a for Criticality Safety Analyses On the HP J-5600 (CMODB) Workstation

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Published: 2001

Total Pages: 101

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This report documents establishment of bias, bias trends and uncertainty for validation of the CSAS25 control module from the SCALE 4.4a computer code system for use in evaluating criticality safety of uranium systems. The 27-group ENDF/B-IV, 44-group ENDF/B-V, and 238-group ENDF/B-V cross-section libraries were used. The criticality validation calculations were performed using over 500 benchmark cases from Volumes II and IV of the ''International Handbook of Evaluated Criticality Safety Benchmark Experiments, '' published by the Nuclear Energy Agency Organization for Economic Cooperation and Development (NEA/OECD). Based on statistical analysis of the calculation results, the bias, bias trends and uncertainty of the benchmark calculations have been established for these benchmark experiments. Numerical methods for applying margins are briefly described, but the determination of appropriate correlating parameter and values for additional margin, applicable to a particular analysis, must be determined as part of process analysis. As such, this document does not specify upper subcritical limits as has been done in the past. A follow-on report will be written to assess the methods for determination of an upper safety limit in more detail, provide comparisons, and recommend a preferred method. Analysts using these results are responsible for exercising sound engineering judgment using strong technical arguments to develop a margin in k{sub eff} or other correlating parameter that is sufficiently large to ensure that conditions (calculated by this method to be subcritical by this margin) will actually be subcritical. Documentation of determination and justification of the appropriate margin in the analyst's evaluation, in conjunction with this report, will constitute the complete Validation Report in accordance with ANSI/ANS-8.1-1998, Section 4.3.6(4).