MCNP criticality validation and bias for LEU systems

The bias in MCP calculations was evaluated for low enriched uranium (LEU) systems typical of N reactor fuel. A formula that includes the bias and its uncertainties is given to ensure that LEU systems are safely subcritical

Comparison of SAND-II and FERRET

A comparison was made of the advantages and disadvantages of two codes, SAND-II and FERRET, for determining the neutron flux spectrum and uncertainty from experimental dosimeter measurements as anticipated in the FFTF Reactor Characterization Program. This comparison involved an examination of the methodology and the operational performance of each code. The merits of each code were identified with respect to theoretical basis, directness of method, solution uniqueness, subjective influences, and sensitivity to various input parameters

Decay heat for the Fast Test Reactor (FTR)

The decay heat for the Fast Test Reactor (FTR) is reevaluated for cooling times up to 30 years. Decay heat from both fission-products and actinides is included. Fission-product decay data used in this evaluation are taken from ENDF/B version IV. Consideration is given to the effects of fuel cycles and loadings and to the importance of fissions in the minor fuels such as /sup 235/U, /sup 240/Pu, and /sup 241/Pu, as well as in the major fuels, /sup 238/U and /sup 239/Pu. Decay-heat uncertainties are evaluated and are typically less than 5 percent for cooling times beyond 50 seconds. The effective energy release per fission for FTR is also reevaluated

Varied applications of a new maximum-likelihood code with complete covariance capability. [FERRET, for data adjustment]

Applications of a new data-adjustment code are given. The method is based on a maximum-likelihood extension of generalized least-squares methods that allow complete covariance descriptions for the input data and the final adjusted data evaluations. The maximum-likelihood approach is used with a generalized log-normal distribution that provides a way to treat problems with large uncertainties and that circumvents the problem of negative values that can occur for physically positive quantities. The computer code, FERRET, is written to enable the user to apply it to a large variety of problems by modifying only the input subroutine. The following applications are discussed: A 75-group a priori damage function is adjusted by as much as a factor of two by use of 14 integral measurements in different reactor spectra. Reactor spectra and dosimeter cross sections are simultaneously adjusted on the basis of both integral measurements and experimental proton-recoil spectra. The simultaneous use of measured reaction rates, measured worths, microscopic measurements, and theoretical models are used to evaluate dosimeter and fission-product cross sections. Applications in the data reduction of neutron cross section measurements and in the evaluation of reactor after-heat are also considered. 6 figures

Finite element basis used in consistent nuclear data evaluation

A method for the consistent evaluation of nuclear cross sections and other data is presented. The method allows the simultaneous inclusion of nuclear model calculations, microscopic and integral measurements, and the results of previously adjusted multigroup cross sections in a consistent evaluation. Complete covariance information is retained throughout the analysis

Activity of TRIGA core components

The activity of TRIGA core components was estimated

Uncertainty analysis for estimates of trapped gas

An uncertainty analysis was made for the amount of trapped gas based on the barometric pressure method for the situation where the method indicates small amounts of gas

SMTAG: A code for the sequential analysis of multiple tag gas releases

The code SMTAG (Sequential and Multiple TAG Analysis) is used to identify breached reactor components that have released tag gas to the reactor cover gas. Gas tags have been used (Figg et al. 1980 and Lambert 1978) to locate failed fuel pins in both the Fast Flux Test Facility (FFTF) and in the Experimental Breeder Reactor (EBR-2). In the FFTF, other reactor components have been tagged as well, including control assemblies and materials test capsules. The SMTAG code has been used extensively in gas tag analysis. This has resulted in several code enhancements and has been beneficial in learning to use the code effectively. Supporting information for each analysis is provided that is valuable in ensuring that a correct identification is obtained. The relative amounts of various components in a mixed sample are obtained, including the amount of residual gas from previous leakers, fission-product release-to-birth factors, and xenon-hangup. Statistical tests and other comparisons can flag bad or inconsistent measurements or problems in the supporting nuclear data base. The formalism for the code is reviewed here in Section 2.0. Details of the code (including descriptions of the main subroutines) are given in Section 3.0. The use of the code is documented in Section 4.0, along with a discussion of a realistic example. The SMTAG code requires a data base that includes the isotopic amounts of each tag properly corrected for burnup, depletion, and production