IBM-704 CODES FOR REACTIVITY STEP CALCULATIONS (RE-126 AND RE-135)

Two codes were written for the IBM-704 to calculate the behavior of a reactor following a step change in reactivity, using one-group. space- independent, zero-power kinetic theory. The reactor is assumed to be running at constant level before the step is made, either at critical or subcritical conditions, with an external source. The code RE-128 assumes all delayed-neutron precursors in equilibrium at the time of the step, while the code RE-135 allows cases with nonequilibrium precursors to be handled. RE-126 can handle the case of zero final reactivity. Both codes are written in FORTRAN language. (C.J.G.

The Mass Tracking System -- Computerized support for MC and A and operations at FCF

As part of Argonne National Laboratory`s Fuel Conditioning Facility (FCF), a computer-based Mass-Tracking (MTG) System has been developed. The MTG System collects, stores, retrieves and processes data on all operations which directly affect the flow of process material through FCF and supports such activities as process modeling, compliance with operating limits (e.g., criticality safety), material control and accountability and operational information services. Its architecture is client/server, with input and output connections to operator`s equipment-control stations on the floor of FCF as well as to dumb terminals and terminal emulators. Its heterogeneous database includes a relational-database manager as well as both binary and ASCII data files. The design of the database, and the software that supports it, is based on a model of discrete accountable items distributed in space and time and constitutes a complete historical record of the material processed in FCF. Although still under development, much of the MTG system has been qualified and is in production use

Argonne National Laboratory Reports

REBUS-3 is a system of programs designed for the fuel-cycle analysis of fast reactors. This new capability is an extension and refinement of the REBUS-3 code system and complies with the standard code practices and interface dataset specifications of the Committee on Computer Code Coordination (CCCC). The new code is hence divorced from the earlier ARC System. In addition, the coding has been designed to enhance code exportability. >Major new capabilities not available in the REBUS-2 code system include a search on burn cycle time to achieve a specified value for the multiplication constant at the end of the burn step; a general non-repetitive fuel-management capability including temporary out-of-core fuel storage, loading of fresh fuel, and subsequent retrieval and reloading of fuel; significantly expanded user input checking; expanded output edits; provision of prestored burnup chains to simplify user input; option of fixed-or free-field BCD input formats; and, choice of finite difference, nodal or spatial flux-synthesis neutronics in one-, two-, or three-dimensions

User's guide for the REBUS-3 fuel cycle analysis capability

REBUS-3 is a system of programs designed for the fuel-cycle analysis of fast reactors. This new capability is an extension and refinement of the REBUS-3 code system and complies with the standard code practices and interface dataset specifications of the Committee on Computer Code Coordination (CCCC). The new code is hence divorced from the earlier ARC System. In addition, the coding has been designed to enhance code exportability. Major new capabilities not available in the REBUS-2 code system include a search on burn cycle time to achieve a specified value for the multiplication constant at the end of the burn step; a general non-repetitive fuel-management capability including temporary out-of-core fuel storage, loading of fresh fuel, and subsequent retrieval and reloading of fuel; significantly expanded user input checking; expanded output edits; provision of prestored burnup chains to simplify user input; option of fixed-or free-field BCD input formats; and, choice of finite difference, nodal or spatial flux-synthesis neutronics in one-, two-, or three-dimensions

CONCEPTUAL DESIGN OF A COUPLED BREEDING SUPERHEATING REACTOR, CBSR

The conceptual design of the Coupled Breeding Superheating Reactor, CBSR, for achieving a positive breeding gain and for producing 65 Mw of electric power is presented. The design combines a steam-cooled fast region and a nonboiling pressurized light-water-cooled thermal region. The advantage offered by this arrangement as compared with that using a solid moderator in the thermal zone is that, if a power excursion occurs, the water will increase the void content and tend to limit the excursion. The total reactor power is 216 Mwt, of which 163 Mwt is used to superheat steam as it passes through the fast regions of the reactor and 53 Mwt is transferred to the pressurized water. For this power split the fast core is 4% subcritical without the reactivity contribution of the thermal region. A breeding ratio of 1.4 is calculated for an oxide-fueled fast and thermal core and a high-density, metal-fueled radial blanket. The steam throttle conditions produced are 75 atm and 453 deg C for an average fast-core power density of 500 Mw/l. The original goal of 565 deg C throttle steam temperature and 1 Mw/l power density was compromised because of the surface temperature limitation of currently available cladding materials. The system does not require a large external power source for producing the steam introduced into the fast core. This is possible through the use of a steam compressor that increases the pressure of a portion of the superheated steam and thus permits its use to generate the required saturated-steam flow rate by vaporizing the feedwater from the steam cycle. The design includes a pressure-balance system that equalizes the static pressure in both the pressurized-water and steam systems. The pressure-balance system provides a means of cooling the steam regions in an emergency by allowing the pressurized water to flash. These features are intended to permit easier startup, operation, and shutdown of the en tire system. A summary of the reactor design characteristics is tabulated. (auth

Improved algorithms for the calculation of resolved resonance cross sections with applications to the structural Doppler effect in fast reactors

Motivated by a need for an economical yet rigorous tool which can address the computation of the structural material Doppler effect, an extremely efficient improved RABANL capability has been developed utilizing the fact that the Doppler broadened line shape functions become essentially identical to the natural line shape functions or Lorentzian limits beyond about 100 Doppler widths from the resonance energy, or when the natural width exceeds about 200 Doppler widths. The computational efficiency has been further enhanced by preprocessing or screening a significant number of selected resonances during library preparation into composition and temperature independent smooth background cross sections. The resonances which are suitable for such pre-processing are those which are either very broad or those which are very weak. The former contribute very little to the Doppler effect and their self-shielding effect can readily be averaged into slowly varying background cross section data, while the latter contribute very little to either the Doppler or to self-shielding effects. To illustrate the accuracy and efficiency of the improved RABANL algorithms and resonance screening techniques, calculations have been performed for two systems, the first with a composition typical of the STF converter region and the second typical of an LMFBR core composition. Excellent agreement has been found for RABANL compared to the reference Monte Carlo solution obtained using the code VIM, and improved results have also been obtained for the narrow resonance approximation in the ultra-fine-group option of MC/sup 2/-2

Conceptual Design of a Coupled Breeding Superheating Reactor, CBSR

The conceptual design of the Coupled Breeding Superheating Reactor, CBSR, for achieving a positive breeding gain and for producing 65 Mw of electric power is presented. The design combines a steam-cooled fast region and a nonboiling pressurized light-water-cooled thermal region. The advantage offered by this arrangement as compared with that using a solid moderator in the thermal zone is that, if a power excursion occurs, the water will increase the void content and tend to limit the excursion. The total reactor power is 216 Mwt, of which 163 Mwt is used to superheat steam as it passes through the fast regions of the reactor and 53 Mwt is transferred to the pressurized water. For this power split the fast core is 4% subcritical without the reactivity contribution of the thermal region. A breeding ratio of 1.4 is calculated for an oxide-fueled fast and thermal core and a high-density, metal-fueled radial blanket. The steam throttle conditions produced are 75 atm and 453 deg C for an average fast-core power density of 500 Mw/l. The original goal of 565 deg C throttle steam temperature and 1 Mw/l power density was compromised because of the surface temperature limitation of currently available cladding materials. The system does not require a large external power source for producing the steam introduced into the fast core. This is possible through the use of a steam compressor that increases the pressure of a portion of the superheated steam and thus permits its use to generate the required saturated-steam flow rate by vaporizing the feedwater from the steam cycle. The design includes a pressure-balance system that equalizes the static pressure in both the pressurized-water and steam systems. The pressure-balance system provides a means of cooling the steam regions in an emergency by allowing the pressurized water to flash. These features are intended to permit easier startup, operation, and shutdown of the en tire system. A summary of the reactor design characteristics is tabulated. (auth

Validation and application of a physics database for fast reactor fuel cycle analysis

An effort has been made to automate the execution of fast reactor fuel cycle analysis, using EBR-II as a demonstration vehicle, and to validate the analysis results for application to the IFR closed fuel cycle demonstration at EBR-II and its fuel cycle facility. This effort has included: (1) the application of the standard ANL depletion codes to perform core-follow analyses for an extensive series of EBR-II runs, (2) incorporation of the EBR-II data into a physics database, (3) development and verification of software to update, maintain and verify the database files, (4) development and validation of fuel cycle models and methodology, (5) development and verification of software which utilizes this physics database to automate the application of the ANL depletion codes, methods and models to perform the core-follow analysis, and (6) validation studies of the ANL depletion codes and of their application in support of anticipated near-term operations in EBR-II and the Fuel Cycle Facility. Results of the validation tests indicate the physics database and associated analysis codes and procedures are adequate to predict required quantities in support of early phases of FCF operations

Argonne National Laboratory Reports

This report describes the current state of the utility subroutine package used with codes being developed by the staff of the Applied Physics Division. The package provides a variety of useful functions for BCD input processing, dynamic core-storage allocation and management, binary I/O and data manipulation. The routines were written to conform to coding standards which facilitate the exchange of programs between different computers