Russell D. Mosteller

MCNP™ Version 5

Abstract The Monte Carlo transport workhorse, MCNP [Los Alamos National Laboratory report LA-13709-M, 2000], is undergoing a massive renovation at Los Alamos National Laboratory (LANL) in support of the Eolus Project of the Advanced Simulation and Computing (ASCI) Program. MCNP 1 Version 5 (V5) (expected to be released to RSICC in Fall 2002) will consist of a major restructuring from FORTRAN-77 (with extensions) to ANSI-standard FORTRAN-90 [American National Standard for Programming Language – Fortran-Extended, ANSI X3. 198-1992, 1992] with support for all of the features available in the present release (MCNP-4C2/4C3). To most users, the look-and-feel of MCNP will not change much except for the improvements (improved graphics, easier installation, better online documentation). For example, even with the major format change, full support for incremental patching will still be provided. In addition to the language and style updates, MCNP V5 will have various new user features. These include improved photon physics, neutral particle radiography, enhancements and additions to variance reduction methods, new source options, improved parallelism support (PVM, MPI, OpenMP), and new nuclear and atomic data libraries.

Benchmark Calculations for the Doppler Coefficient of Reactivity

This paper reports on the Doppler coefficient of reactivity that is a crucial parameter in the evaluation of several transients in light water reactors (LWRs). It is relatively small in magnitude and cannot be measured directly in operating reactors. Doppler coefficients are presented for slightly idealized pressurized water reactor pin cells. These coefficients were calculated with the MCNP-3A continuous-energy Monte Carlo code using data taken directly from the ENDF/B-V nuclear data library. This combination represents the most rigorous analytical tool and the best nuclear data available. Consequently, these results comprise a set of numerical benchmarks that may be used to evaluate the accuracy of LWR lattice physics codes in predicting Doppler behavior at operating conditions. An example of one such evaluation, using the CELL-2 code, is included.

Monte Carlo calculations for recriticality during the reflood phase of a severe accident in a boiling water reactor

The possibility of recriticality during the reflood phase of a severe accident in a boiling water reactor (BWR) is investigated. In addition, the fraction of control-rod material that must be retained to prevent recriticality is determined for several different geometries. Detailed calculations for a representative portion of a BWR core are performed with the Monte Carlo MCNP code to evaluate the neutronic response of the damaged core under a number of different scenarios. Retention of <10% of the {sup 10}B that is initially present in the control rods may be sufficient to prevent recriticality. The reactivity effect of the retained {sup 10}B is sensitive to the porosity and homogeneity of the eutectic formed by the control rod materials as well as to the degree to which that eutectic is dispersed inside the fuel bundles.

Data testing of ENDF/B-VI with MCNP: Critical experiments, thermal-reactor lattices, and time-of-flight measurements

The United States evaluated nuclear database, ENDF/B, is organized and implemented by the Cross Section Evaluation Working Group (CSEWG), which is a cooperative industrial/governmental activity involving, at its peak, some 20 different organizations. Since its inception in 1966, CSEWG systematically improved the evaluated database as new experimental and theoretical information became available, periodically issuing new versions of the ENDF/B library. The ENDF/B-VI file initially was issued in 1989-1990 and has been followed by three updated releases, with a fourth to follow soon. The purpose of this paper is to review the status of the ENDF/B-VI data file and to describe recent data testing results obtained with the Monte Carlo N-Particle (MCNP) code.

Comparison of results from the MCNP criticality validation suite using ENDF/B-VI and preliminary ENDF/B-VII nuclear data

The MCNP Criticality Validation Suite is a collection of 31 benchmarks taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments. MCNP5 calculations clearly demonstrate that, overall, nuclear data for a preliminary version of ENDF/B‐VII produce better agreement with the benchmarks in the suite than do corresponding data from ENDF/B‐VI. Additional calculations identify areas where improvements in the data still are needed.

Analysis of the first three Zeus critical experiments

Abstract The Zeus experiments have been designed to test the adequacy of 235U cross sections in the intermediate-energy range. Detailed models of the three Zeus critical experiments are developed for the MCNP Monte Carlo code, and calculated results, based on cross sections derived from ENDF/B-V and ENDF/B-VI, are presented and assessed. A series of modeling simplifications then is described that transforms the detailed representations into benchmark configurations, and the reactivity impact of each of those simplifications is assessed.

Analysis of a pressurized water reactor natural circulation transient at beginning of life

A pressurized water reactor (PWR) with a positive moderator temperature coefficient of reactivity is potentially susceptible to a severe overheating transient. This study identifies a scenario in which such a transient could occur and is similar in some respects to the accident at Chernobyl Unit 4. The scenario so identified is a natural circulation test at beginning of life under the assumption that all scrams are disabled. The results obtained demonstrate that a runaway power excursion does not occur and that the domestically designed PWR that was analyzed displays inherently safe behavior for the chosen scenario. The analysis is performed using two codes in tandem over three sequential stages of the analysis.