ENDF/B-V and ENDF/B-VI results for UO{sub 2} lattice benchmark problems using MCNP

Calculations for the ANS UO{sub 2} lattice benchmark have been performed with the MCNP Monte Carlo code and its ENDF/B-V and ENDF/B-VI continuous-energy libraries. The ENDF/B-V library produces significantly better agreement with the benchmark value for k{sub eff} than do the ENDF/B-VI libraries. However, the pin power distributions are essentially the same irrespective of the library

Validation of NESTLE against static reactor benchmark problems

The NESTLE advanced modal code was developed at North Carolina State University with support from Los Alamos National Laboratory and Idaho National Engineering Laboratory. It recently has been benchmarked successfully against measured data from pressurized water reactors (PWRs). However, NESTLE`s geometric capabilities are very flexible, and it can be applied to a variety of other types of reactors. This study presents comparisons of NESTLE results with those from other codes for static benchmark problems for PWRs, boiling water reactors (BWRs), high-temperature gas-cooled reactors (HTGRs) and CANDU heavy- water reactors (HWRs)

Development of an automated core model for nuclear reactors

This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input

HELIOS calculations for UO{sub 2} lattice benchmarks

Calculations for the ANS UO{sub 2} lattice benchmark have been performed with the HELIOS lattice-physics code and six of its cross-section libraries derived from ENDF/B-VI Release 3. The results obtained from these comparisons suggest that further refinement may be needed to the cross sections for {sup 238}U. They also suggest that different group structures among the libraries produce a small but consistent reactivity bias

Benchmarking of NESTLE against measured PWR data at beginning of life

The NESTLE advanced nodal code was developed at North Carolina State University with support from Los Alamos National Laboratory and Idaho National Engineering Laboratory. This paper presents the first comparisons of NESTLE predictions with measured data from pressurized water reactors (PWRs). Specifically, NESTLE predictions for critical soluble boron concentrations and isothermal temperature coefficients of reactivity (ITCs) are compared with beginning-of-life (BoL) measurements from four PWRs. All of those measurements were made at hot-zero-power (HZP) conditions prior to ascension to power

Static benchmarking of the NESTLE advanced nodal code

Results from the NESTLE advanced nodal code are presented for multidimensional numerical benchmarks representing four different types of reactors, and predictions from NESTLE are compared with measured data from pressurized water reactors (PWRs). The numerical benchmarks include cases representative of PWRs, boiling water reactors (BWRs), CANDU heavy water reactors (HWRs), and high-temperature gas-cooled reactors (HTGRs). The measured PWR data include critical soluble boron concentrations and isothermal temperature coefficients of reactivity. The results demonstrate that NESTLE correctly solves the multigroup diffusion equations for both Cartesian and hexagonal geometries, that it reliably calculates k{sub eff} and reactivity coefficients for PWRs, and that--subsequent to the incorporation of additional thermal-hydraulic models--it will be able to perform accurate calculations for the corresponding parameters in BWRs, HWRs, and HTGRs as well

Impact of MCNP unresolved resonance probability-table treatment on {sup 233}U benchmarks

Previous versions of the MCNP Monte Carlo code, up through and including MCNP4B, have not accurately modeled neutron self-shielding effects in the unresolved resonance energy region. Recently, a probability-table treatment has been incorporated into an intermediate version called MCNP4XS, and compatible continuous-energy cross-section libraries have been developed for 27 different isotopes. Preliminary results for a variety of uranium and plutonium benchmarks have been presented previously, and this paper extends those results to include several {sup 233}U benchmarks. The objective of the current study is to assess the reactivity impact of the probability-table treatment on {sup 233}U systems

Two-dimensional benchmark calculations for PNL-30 through PNL-35

Interest in critical experiments with lattices of mixed-oxide (MOX) fuel pins has been revived by the possibility that light water reactors will be used for disposition of weapons-grade plutonium. A series of six experiments with MOX lattices, designated PNL-30 through PNL-35, was performed at Pacific Northwest Laboratories in 1975 and 1976, and a set of benchmark specifications for these experiments subsequently was adopted by the Cross Section Evaluation Working Group (CSEWG). Although there appear to be some problems with these experiments, they remain the only CSEWG benchmarks for MOX lattices. The number of fuel pins in these experiments is relatively low, corresponding to fewer than 4 typical pressurized-water-reactor fuel assemblies. Accordingly, they are more appropriate as benchmarks for lattice-physics codes than for reactor-core simulator codes. Unfortunately, the CSEWG specifications retain the full three-dimensional (3D) detail of the experiments, while lattice-physics codes almost universally are limited to two dimensions (2D). This paper proposes an extension of the benchmark specifications to include a 2D model, and it justifies that extension by comparing results from the MCNP Monte Carlo code for the 2D and 3D specifications

Benchmarking of MCNP against B&W LRC Core XI critical experiments

The MCNP Monte Carlo code and its ENDF/B-V continuous-energy cross- section library previously has been benchmarked against a variety of critical experiments, and that benchmarking recently has been extended to include its ENDF/B-VI continuous-energy cross-section library and additional critical experiments. This study further extends the benchmarking of MCNP and its two continuous-energy libraries to 17 large-scale mockup experiments that closely resemble the core of a pressurized water reactor (PWR). The experiments were performed at Babcock & Wilcox`s Lynchburg Research Center in 1970 and 1971. The series was designated as Core XI, and the individual experiments were characterized as different ``loadings.`` The experiments were performed inside a large aluminum tank that contained borated water. The water height for each loading was exactly 145 cm, and the soluble boron concentration in the water was adjusted until the configuration was slightly supercritical, with a value of 1.0007 for k{sub eff}. Pin-by-pin power distributions were measured for several of the loadings

Development of a standard for the moderator temperature coefficient of reactivity in water-moderated power reactors

The moderator temperature coefficient of reactivity (MTC) is an important parameter in safety analyses for thermal reactors. A positive MTC can exacerbate the severity of heatup transients, while a negative MTC can worsen the severity of cooldown transients. In particular, a strongly negative MTC is the major determinate of the severity of the steamline-break accident for pressurized water reactors (PWRs). Conversely, positive and negative MTCs can mitigate the severity of cooldown and heatup transients, respectively. Consequently, the accurate measurement and prediction of MTCs is an important factor in demonstrating that power reactors can be operated safely. This document discusses the development of a standard for measuring MTC