Fast Reactor Analysis Research Articles

Dominant Delayed Neutron Precursors to Model Reactivity Predictions for Multiple Fissioning Nuclides

The product of cumulative yield and probability of neutron emission is used to assess the relative importance of known delayed neutron precursors. Thirteen precursors are consistently dominant. Nonlinear fits to experimental delayed neutron decay data distinguish the decay constants of the three longest-lived dominant precursors: 87Br, 137I, and 88Br. Sensitivity calculations based on a six- to seven- group transformation lead to a proposed seven-group formulation in which the group decay constants are those of dominant precursors: 87Br, 137I, 88Br, 93Rb, 139I, 91Br, and 96Rb. An alternative six-group formulation is obtained by using the mean of the 137I and 88Br decay constants for group 2. The use of the suggested dominant precursor decay constants improves the goodness of fit to experimental data compared to that obtained from nonlinear least squares in which both group yields and decay constants are determined empirically. Reactivity worth and transient analyses confirm that the positive reactivity scale is preserved in the transformation. A known bias in the negative reactivity scale is eliminated by forcing the half-life of the longest-lived group to be the 55.9-s half-life of 87Br. The proposed use of dominant precursor decay constants offers significant simplifications in data analysis and the analysis of fast, epithermal, and thermal reactors with multiple fissioning nuclides.

The Development of a Preprocessor to Generate Input for Monte Carlo Codes and Its Application to Fast Reactor Analysis

For many calculations in Reactor Physics, Monte Carlo methods have significant advantages over standard deterministic techniques. However, because very accurate geometric models can be generated, input data to Monte Carlo codes is difficult, tedious and extremely time consuming to prepare. This paper describes a preprocessor which reduces considerably the size and complexity of user generated data. In fact only dimensional and composition information is required by the preprocessor for it to produce the code input file and it is believed that this is a significant development in facilitating Monte Carlo data preparation.After a description of the preprocessor an indication of the reduction in effort required to generate the Monte Carlo input for the heterogeneous pin structure of MASURCA by using the preprocessor is given followed by comparisons of calculated and experimental keff values. Finally future developments and applications of the preprocessor are described.In this paper only applications to Fast Reactor Physics will be considered. However the approach is quite general and could be extended to Thermal studies.

The Development of a Preprocessor to Generate Input for Monte Carlo Codes and Its Application to Fast Reactor Analysis.

For many calculations in Reactor Physics, Monte Carlo methods have significant advantages over standard deterministic techniques. However, because very accurate geometric models can be generated, input data to Monte Carlo codes is difficult, tedious and extremely time consuming to prepare. This paper describes a preprocessor which reduces considerably the size and complexity of user generated data. In fact only dimensional and composition information is required by the preprocessor for it to produce the code input file and it is believed that this is a significant development in facilitating Monte Carlo data preparation. After a description of the preprocessor an indication of the reduction in effort required to generate the Monte Carlo input for the heterogeneous pin structure of MASURCA by using the preprocessor is given followed by comparisons of calculated and experimental keff values. Finally future developments and applications of the preprocessor are described. In this paper only applications to Fast Reactor Physics will be considered. However the approach is quite general and could be extended to Thermal studies.

Three-Dimensional Transient Analysis of Fast Reactors with Improved Coarse Mesh Method.

To estimate fast reactor core transient behavior in three-dimensional geometry, we have applied the improved coarse mesh method to time dependent problems. The improved quasistatic approximation is used with the feedback of fuel temperature change and coolant temperature change. The shape function describing flux distribution is calculated using effective cross sections which are functions of mesh intervals and the amplitude function. To allow large axial coarse meshes with control rod partially inserted, a bilinearly weighting method is used to calculate homogenized cross sections. Using these methods KICOM code has been developed. The validity of this code has been checked by comparing with a two-dimensional transient code FX2-TH for the case when a control rod at core center is withdrawn. The KICOM code has been applied to calculate the power change of a prototype fast reactor and a large fast reactor for the case when a core center rod or an off-center rod is withdrawn. For the prototype reactor, the ...

Three-Dimensional Transient Analysis of Fast Reactors with Improved Coarse Mesh Method

Three-Dimensional Transient Analysis of Fast Reactors with Improved Coarse Mesh Method

Ultra-miniature chip components change mounting methods : Isamu Ishiguro. JEE (Japan), 36 (November 1992)

Thermal–hydraulic analysis of sodium-cooled fast reactors (SFRs) test subassemblies has been performed employing subchannel analysis method. In order to improve the accuracy of the calculation, recent empirical correlations of pressure drop coefficient and inter-channel mixing for evaluation of coolant flow in wire-wrapped fuel subassemblies have been incorporated into the subchannel analysis model. To confirm the prediction capability of the model, calculations and comparisons with available experimental data of 19- and 37-pin subassemblies have been performed. A comparison with previous results of a 127-pin subassembly has also been performed. The results show that empirical correlations which consider three subchannel types separately are capable of predicting the coolant temperature distribution at bundle outlet best among models. Moreover it has been found that the agreement between calculation result and experimental data of outlet temperature distribution improves when bundle hot dimensions are employed. In addition to the conventional flow resistance models, performance of a method of directly varying coolant flow areas to model geometrical disturbance due to the presence of wire spacers has been evaluated. It has been found that the method is capable of simulating crossflows between adjacent subchannels. All the results show reasonably good agreement with the reference data, in which the root mean square (RMS) differences are within 10% for all cases analyzed in this study.

Assessment of Life of Inner Vessel for a Pool-Type Fast Breeder Reactor Based on Inelastic and Viscoplastic Deformations

Towards assessing the creep-fatigue damage for the inner vessel of a pool-type fast reactor, 3-D elastic and axisymmetric inelastic analysis with classical kinematic hardening model, ORNL material model and more realistic CHABOCHE visco-plastic model have been performed. The material properties of stainless steel type 316 LN at various temperatures are used in the analysis. Using the results of this analysis, the creep-fatigue damage and ratcheting are assessed as per the rules of RCC-MR and results have been compared for the various constitutive models used. Inelastic analysis has been performed only at highly stressed local regions. Computations with the ORNL material model is found to be satisfactory in terms of conservatism and computer cost involved, and hence recommended for use in the design stage. Further, conservatism in estimating the creep fatigue-damage of the component associated with the elastic route of RCC-MR has been quantified. The inelastic and viscoplastic analysis tool has been validated using experimental benchmark problems.

Dynamics of Fuel Vapor Pressure Buildup in Voided LMFBR Cores During the Transient Heating and Its Effect on the Energy Release in a Core Disruptive Accident

The core disruptive accident (CDA) analysis of liquid-metal fast breeder reactors is often performed using the saturated vapor pressure equation of state for the fuel. However, during the transient...

Dynamic response of an lmfbr deck structure under slug impact

An important concern in the safety analysis of liquid-metal fast breeder reactors (LMFBR) is the dynamic response of the primary containment to the impact of the fluid occurring during a hypothetical core disruptive accident (HCDA). It is possible that a substantial part of the kinetic energy gained by the coolant from the energy released by the core is transferred to the deck structure which, however, should retain its integrity in order to prevent leakage of radioactive material to the environment. The dynamic response of the deck structure of a pool-type plant of commercial size is analyzed in a generic study using the program ADINA. The assumed energy release during the hypothetical accident represents a work capacity of 0.5 GJ. The resulting slug impact loading applied to the deck structure is based on a fluid-structure interaction analysis which assumes axial symmetry. As a reference deck, a box-type structure of annular form with concrete fill is chosen.

Lumped Fission Product Neutron Cross Sections Based on ENDF/B-V for Fast Reactor Analysis

The development and evaluation of a lumped fission product neutron cross-section library based on ENDF/B-V data suitable for fast reactor applications have been completed. Both one- and two-lump models have been investigated in detail. Fission product inventories at various burnup steps were calculated by the EPRI-CINDER-2 code and used as weighting functions for lumping. This paper addresses several important issues related to the lumped data including the relative merits of the two models, the dependence on burnup history, the influence of fuel composition and neutron spectrum, the impact of various data bases, the application of the lumped data, the effect of the scattering matrix, and finally the impact on the fission product reactivity worth in a fast reactor system. Although the data and results contained in this paper are specifically related to a particular mixed-oxide core design, they have general validity and application to other similar liquid-metal fast breeder reactor designs.

Authors

Authors

Investigation of Local Cooling Disturbances in an In-Pile Sodium Loop in the BR2

The fuel element behavior under local off-normal cooling conditions and the possible pin-to-pin failure propagation are of special interest in the safety analysis of liquid-metal fast breeder reactors. In a program called Mol 7C, Kernforschungszentrum Karlsruhe and the Centre d’Etude de l’Energie Nucléaire/Studiecentrum voor Kernenergie at Mol are performing related experiments in a sodium loop in the BR2 reactor. The test section contains a 37-pin bundle of UO2 fuel with an artificial local blockage involving the 6 inner pins and parts of the pins of the second row (30 rods are active fuel pins, and 7 are stainless-steel dummy pins). After a pre-irradiation of some days, the transient test phase is initiated by interrupting the cooling in the local blockage region. The first and second of three planned experiments demonstrate that the cooling of the whole fuel bundle was not jeopardized, although several pins have failed. The performance of the third experiment had to be postponed for the end of 1979 due to the exchange of the beryllium matrix of the BR2 reactor during 1979.

Boiling Liquid-Metal Heat Transfer

"Boiling Liquid-Metal Heat Transfer." Nuclear Technology, 43(3), pp. 394–395 Additional informationNotes on contributorsAdrian M. Tentner(PhD, nuclear engineering, University of Cincinnati) is a nuclear engineer at Argonne National Laboratory in the Reactor Analysis and Safety Division. His experience includes eight years of research in transient hydraulics and heat transfer, two-phase flow, and numerical problems arising in nuclear reactor safety analysis.He is currently engaged in the analysis and analytical modeling of postaccident fuel-coolant interactions inliquid-metal fast breeder reactors.

Authors

Authors

Benchmark Analysis of Liquid-Metal Fast Breeder Reactor Nuclear Design Methods

The Large Core Code Evaluation Working Group has been organized for the prime aim of testing and validating neutronics codes and methods for use in the analysis of large fast reactors. For the initial problem, static neutronics calculations were performed on a “representative” 1200-MW(e) liquid-metal fast breeder reactor in a two-dimensional radial geometry. The initial results representing the cooperative effort of the participants are summarized. Various codes and methods using few-group diffusion theory are intercompared. The calculational methods are in substantial agreement provided the same multigroup constants are used. This effort is to be further pursued and broadened to investigate such specific effects as sodium voiding.

A Note on Turbulent-Laminar Transition for Flow in Triangular Rod Bundles

The Ramm-Johannsen correlation of eddy diffusivity for momentum is used to compute the onset and growth of flow laminarization in triangular rod bundles. The laminarization is indicated by Reynolds numbers based on the bulk flow condition in the coolant channel. Simple correlations that qualitatively estimate the transition Reynolds numbers are developed for applications in the safety analysis of liquid-metal fast breeder reactors.

Benchmark Testing Using ENDF/B-III and -IV

A series of benchmark calculations of critical experiments has been performed to assess the effects that recent changes in the ENDF/B data files have had on calculated liquid-metal fast breeder reactor parameters. Three well-documented critical assemblies were studied using standard methods of fast reactor analysis (two-dimensional multigroup diffusion theory) with both ENDF/B-III and -IV. A review of the changes in the principal cross sections incorporated in the latest evaluation has been made and was used to interpret the changes in calculated integral parameters. Some of the principal cross-section modifications included in ENDF/B-IV were: decreasing (≍2%), (≍1 to 1.5%), and (≍3 to 4%) and increasing (≍1 to 1.5%). In general, improved agreement between measured and calculated integral parameters has been obtained with the new ENDF/B evaluation. The overprediction of 28c/49 ƒ is reduced slightly (≍1.5%); the central worth calculation-to-experiment discrepancy is generally reduced (≍5 to 15%); and little change occurs in the calculated eigenvalues that remain ≍1.0 to 1.5% δk low.

The Importance of the Uranium-238 Inelastic Transfer Matrix for Fast-Reactor Analysis

The 238U inelastic scattering was modified to reflect the recently recommended larger values for the 45-keV (2+) and 146-keV (4+) levels. The partial inelastic cross sections for the other levels given in ENDF/B-IV between 0.8 and 3.7MeV were accordingly decreased such as to keep the total inelastic cross section unchanged. To investigate the consequences of these modifications three fast-neutron assemblies have been analyzed. The above modifications result in a general improvement between calculated and measured fast-reactor parameters.

Vapor Pressure Measurements of Oxide Fuel Between 3000 and 5000 K Using Laser Heating

Safety analysis of fast reactors requires knowledge of the vapor pressure of the fuel material under the condition of rapid heating in the temperature region from 3000 K upwards, where up to now no measured data have been available.An experimental technique to determine the vapor pressure of refractory fuel materials in the temperature range between 3000 and 5000 K is based on laser beam heating of the specimen surface for ∼1 msec and measurement of both the recoil momentum of the specimen and the mass and momentum carried away by the vapor jet flowing into a vacuum. The determination of the vapor pressure requires application of gas dynamics and development of a suitable mode. The application limits of such a measuring technique using laser heating and the reliability of the vapor pressure data thereby determined have been assessed.

Flux Synthesis Calculations for Fast Reactors

A number of calculations have been performed to investigate the use of spatial flux synthesis methods in fast reactor analysis. Guidelines are discussed for choosing appropriate expansion functions. The use of a continuous multichannel synthesis formulation is illustrated by a sodium voiding calculation.