Few group collapsing of covariance matrix data based on a conservation principle

A new algorithm for a rigorous collapsing of covariance data is proposed, derived, implemented, and tested. The method is based on a conservation principle that allows preserving at a broad energy group structure the uncertainty calculated in a fine group energy structure for a specific integral parameter, using as weights the associated sensitivity coefficients

Reactor benchmarks and integral data testing and feedback into ENDF/B-VI

The role of integral data testing and its feedback into the ENDF/B evaluated nuclear data files are reviewed. The use of the CSEWG reactor benchmarks in the data testing process is discussed and selected results based on ENDF/B Version VI data are presented. Finally, recommendations are given to improve the implementation in future integral data testing of ENDF/B

Analysis of three idealized reactor configurations: plate, pin, and homogeneous. [LMFBR]

Detailed Monte Carlo calculations have been performed for three distinct configurations of an idealized fast critical assembly. This idealized assembly was based on the LMFBR benchmark critical assembly ZPR-6/7. In the first configuration, the entire core was loaded with the plate unit cell of ZPR-6/7. In the second configuration, the entire core was loaded with the ZPR sodium-filled pin calandria. The actual ZPR pin calandria are loaded with mixed (U,Pu) oxide pins which closely match the composition of the ZPR-6/7 plate unit cell. For the present study, slight adjustments were made in the atom concentrations and the length of the pin calandria in order to make the core boundaries and average composition for the pin-cell configuration identical to those of the plate-cell configuration. In the third configuration, the core was homogeneous, again with identical core boundaries and average composition as the plate and pin configurations

The role of integral experiments and nuclear cross section evaluations in space nuclear reactor design

The importance of the nuclear and neutronic properties of candidate space reactor materials to the design process has been acknowledged as has been the use of benchmark reactor physics experiments to verify and qualify analytical tools used in design, safety, and performance evaluation. Since June 1966, the Cross Section Evaluation Working Group (CSEWG) has acted as an interagency forum for the assessment and evaluation of nuclear reaction data used in the nuclear design process. CSEWG data testing has involved the specification and calculation of benchmark experiments which are used widely for commercial reactor design and safety analysis. These benchmark experiments preceded the issuance oflthe industry standards for acceptance, but the benchmarks exceed the minimum acceptance criteria for such data. Thus, a starting place has been provided in assuring the accuracy and uncertainty of nuclear data important to space reactor applications. (FI

Derivation of criticality safety benchmarks from ZPR fast critical assemblies

Scores of critical assemblies were constructed, over a period of about three decades, at the Argonne National Laboratory ZPR-3, ZPR-6, ZPR-9, and ZPPR fast critical assembly facilities. Most of the assemblies were mockups of various liquid-metal fast breeder reactor designs. These tended to be complex, containing, for example, mockups of control rods and control rod positions. Some assemblies, however, were `physics benchmarks`. These relatively `clean` assemblies had uniform compositions and simple geometry and were designed to test fast reactor physics data and methods. Assemblies in this last category are well suited to form the basis for new criticality safety benchmarks. The purpose of this paper is to present an overview of some of these benchmark candidates and to describe the strategy being used to create the benchmarks

Lessons learned from applying VIM to fast reactor critical experiments

VIM is a continuous energy Monte Carlo code first developed around 1970 for the analysis of plate-type, fast-neutron, zero-power critical assemblies. In most respects, VIM is functionally equivalent to the MCNP code but it has two features that make uniquely suited to the analysis of fast reactor critical experiments: (1) the plate lattice geometry option, which allows efficient description of and neutron tracking in the assembly geometry, and (2) a statistical treatment of neutron cross section data in the unresolved resonance range. Since its inception, VIM`s capabilities have expanded to include numerous features, such as thermal neutron cross sections, photon cross sections, and combinatorial and other geometry options, that have allowed its use in a wide range of neutral-particle transport problems. The earliest validation work at Argonne National Laboratory (ANL) focused on the validation of VIM itself. This work showed that, in order for VIM to be a ``rigorous`` tool, extreme detail in the pointwise Monte Carlo libraries was needed, and the required detail was added. The emphasis soon shifted to validating models, methods, data and codes against VIM. Most of this work was done in the context of analyzing critical experiments in zero power reactor (ZPR) assemblies. The purpose of this paper is to present some of the lessons learned from using VIM in ZPR analysis work. This involves such areas as uncovering problems in deterministic methods and models, pitfalls in using Monte Carlo codes, and improving predictions. The numerical illustrations included here were taken from the extensive documentation cited as references

Proceedings of the NEACRP/IAEA Specialists meeting on the international comparison calculation of a large sodium-cooled fast breeder reactor at Argonne National Laboratory on February 7-9, 1978

The results of an international comparison calculation of a large (1250 MWe) LMFBR benchmark model are presented and discussed. Eight reactor configurations were calculated. Parameters included with the comparison were: eigenvalue, k/sub infinity/, neutron balance data, breeding reaction rate ratios, reactivity worths, central control rod worth, regional sodium void reactivity, core Doppler and effective delayed neutron fraction. Ten countries participated in the comparison, and sixteen solutions were contributed. The discussion focuses on the variation in parameter values, the degree of consistency among the various parameters and solutions, and the identification of unexpected results. The results are displayed and discussed both by individual participants and by groupings of participants (e.g., results from adjusted data sets versus non-adjusted data sets)

Nuclear data needs within the US Nuclear Criticality Safety Program

This paper will present the nuclear data needs currently identified within the US Nuclear Criticality Safety Program (NCSP). It will identify the priority data needs; it will describe the process of prioritizing those needs; and it will provide brief examples of recent data advances which have successfully addressed some of the priority criticality safety data needs

Recent BeO-reflector-controlled reactor experiments in ZPPR

Integral reactor physics measurements were performed on a BeO-reflected fast reactor assembly in the ZPPR facility during January and February of 1985. The measurements emphasized power distributions and reflector control worths in two different critical states. The measurements have been analyzed using three-dimensional deterministic and Monte Carlo methods and the ENDF/B-V.2 nuclear data library. Together the measurements and analyses form a modern, reliable, benchmark data set for testing calculational methods that will be used in predicting some of the design parameters for future space reactors