Sedat Goluoglu

Monte Carlo Criticality Methods and Analysis Capabilities in SCALE

Abstract This paper describes the Monte Carlo codes KENO V.a and KENO-VI in SCALE that are primarily used to calculate multiplication factors and flux distributions of fissile systems. Both codes allow explicit geometric representation of the target systems and are used internationally for safety analyses involving fissile materials. KENO V.a has limiting geometric rules such as no intersections and no rotations. These limitations make KENO V.a execute very efficiently and run very fast. On the other hand, KENO-VI allows very complex geometric modeling. Both KENO codes can utilize either continuous-energy or multigroup cross-section data and have been thoroughly verified and validated with ENDF libraries through ENDF/B-VII.0, which has been first distributed with SCALE 6. Development of the Monte Carlo solution technique and solution methodology as applied in both KENO codes is explained in this paper. Available options and proper application of the options and techniques are also discussed. Finally, performance of the codes is demonstrated using published benchmark problems.

A deterministic method for transient, three-dimensional neutron transport

A deterministic method for solving the time-dependent, three-dimensional Boltzmam transport equation with explicit representation of delayed neutrons has been developed and evaluated. The methodology used in this study for the time variable of the neutron flux is known as the improved quasi-static (IQS) method. The position, energy, and angle-dependent neutron flux is computed deterministically by using the three-dimensional discrete ordinates code TORT. This paper briefly describes the methodology and selected results. The code developed at the University of Tennessee based on this methodology is called TDTORT. TDTORT can be used to model transients involving voided and/or strongly absorbing regions that require transport theory for accuracy. This code can also be used to model either small high-leakage systems, such as space reactors, or asymmetric control rod movements. TDTORT can model step, ramp, step followed by another step, and step followed by ramp type perturbations. It can also model columnwise rod movement can also be modeled. A special case of columnwise rod movement in a three-dimensional model of a boiling water reactor (BWR) with simple adiabatic feedback is also included. TDTORT is verified through several transient one-dimensional, two-dimensional, and three-dimensional benchmark problems. The results show that the transport methodology and corresponding code developed in this work have sufficient accuracy and speed for computing the dynamic behavior of complex multidimensional neutronic systems.

Criticality Analysis of MOX and LEU Assemblies for Transport and Storage at the Balakovo Nuclear Power Plant

Abstract Criticality of low-enriched-uranium (LEU) and mixed-oxide (MOX) assemblies at the VVER-1000-type Balakovo nuclear power plant is investigated. Effective multiplication factors for fresh fuel assemblies on the railroad platform, fresh fuel assemblies in the within-plant fuel transportation vehicle, and fresh fuel assemblies in the spent fuel storage pool are calculated. If there is no absorber between the units, the configurations with all MOX assemblies result in higher effective multiplication factors than the configurations with all LEU assemblies when the system is dry. When the systems are flooded, the configurations with all LEU assemblies result in higher effective multiplication factors. For normal operating conditions, effective multiplication factors for all configurations are below the presumed upper subcritical limit of 0.95. For an accident condition of a fully loaded within-plant fuel transportation vehicle that is “flooded” with low-density water (possibly from a fire suppression system), the presumed upper subcritical limit is exceeded by configurations containing either LEU or a combination of LEU and MOX assemblies.

Criticality Safety Evaluation of Shutdown Diffusion Cascade Coolers

A criticality safety study of diffusion cascade coolers in a shutdown state is presented. The coolers represent six typical cascade coolers at a gaseous diffusion plant with accumulated deposits of UO 2 F 2 . The study involves k eff calculations for the coolers with various distributions of UO 2 F 2 , which are assumed as part of several hypothetical accident scenarios. The results show that at least two independent failures must occur in order to have a criticality. Additionally, the distributions chosen represent the upper bounds for k eff . Individual results show that the k eff values for the cascade coolers designed for 80 and 97% enriched UF 6 with deposit amounts <2.409 and 2.185 kg, respectively, will not exceed 0.9 for the accident scenarios modeled. All other coolers require shell-side flooding with H 2 O in order to cause a criticality, which is possible only if two or more independent failures occur.