An adiabatic kinetics model has been developed for a BWR core simulator AETNA. A few-energy-group analytic polynomial nodal diffusion method consistent with its steady-state solution has been adopted for spatial solution. The frequency transform method has been introduced to reduce computing time during the transient. The AETNA solution efficacy has been evaluated through comparison of the results with other codes using established benchmark problems. In particular, AETNA has shown a good agreement with QUANDRY, i.e., almost the same accuracy as other codes with regard to power or power density, peak time and fuel temperature. The computing time is reduced to less than one-ninth in no feedback cases and less than one-third in a Doppler feedback case when applying the frequency transform method.
In boiling water reactor (BWR) cores, the radial void distribution in fuel bundles is thought to deviate from uniform distribution. The effect of heterogeneity in the subchannel void fraction distribution, caused by the presence of Gd-poisoned and cold surfaces as well as control blades on BWR lattice physics parameters, has been evaluated. The cross-sectional void distributions in each axial plane of a fuel bundle are calculated using the subchannel thermal hydraulics code COBRAG. The rod power distributions to be fed to COBRAG are calculated using either the Monte Carlo code MCNP4C or the fuel lattice code TGBLA. Iterative methods consisting of COBRAG and MCNP4C (or TGBLA) are established. A set of test cases was generated for a typical BWR 8 Â 8 fuel bundle. The results of the coupled MCNP4C and COBRAG method reveal that both Gd rods and control blade increase in worth due to the void heterogeneity, showing a maximum decrease in K 1 of $0:7%Ák=k 1 and $1:5%Ák=k 1 , respectively, at moderator density conditions equivalent to $40% void fraction. With the capability of the coupled TGBLA and COBRAG method to deplete fuel bundles, the acceleration of Gd depletion was evaluated. The impacts on the burnup characteristics of k 1 reached AE0:6%Ák 1 at maximum.
TRACG code, coupling a three-dimensional neutron kinetics model for the reactor core with thermal-hydraulics based on two-fluid conservation equations, is a best-estimate (BE) code for BWRs to realistically simulate their transient and accidental behaviors. TRACG05 is the latest version and was originally developed to analyze Reactivity Initiated Accident (RIA). TRACG05 incorporates the same neutronics model of the latest core simulator with a three-group analytic-polynomial nodal expansion method. In addition to application to RIA safety analyses, TRACG05 has been planned to apply to safety analyses for Anticipated Operational Occurrences (AOOs) in BWRs by using a Best Estimate Plus Uncertainty (BEPU) methodology. To apply BEPU with TRACG05 to BWR AOOs, validations must be performed to evaluate the uncertainties of models relevant to important phenomena by comparing with appropriate test results for BWR AOOs. At first, a PIRT (Phenomena Identification and Ranking Table) was developed for each event scenario in AOOs to identify relevant physical processes and to determine their relative importance. According to the PIRT, an assessment matrix was established for separate effects tests (SETs), component effects tests (CETs), integral effects tests (IETs), and integral BWR plant start-up tests. The assessment matrix related the important phenomena to the test database, which was confirmed that all the important phenomena were covered by all tests specified in the matrix. According to the assessment matrix, comparison analyses have been specified to perform systematic and comprehensive validations of TRACG05 applicability to AOOs. The comparison analyses were done as the integrated code system with the up-stream reactor core design codes, therefore higher quality was enabled to evaluate the safety parameters. As the result, the uncertainties of important models in TRACG05 were determined so as to enable BEPU approaches for AOO safety issues. Here, as a SET, comparisons between TRACG05 and experimental data of void fraction in a bundle simulating an actual fuel bundle, which is one of the most important models in the application of TRACG05 to AOO analyses are shown. In addition, as pressurization event in AOOs, comparisons between TRACG05 and experimental data of Peach Bottom 2 Turbine Trip Test, which is one of integral tests for a BWR plant, are shown. This is the only test showing large neutron flux increase and strong coupling of neutron kinetics and thermal-hydraulics in the core due to void and Doppler feedbacks. Furthermore, a sensitivity analysis regarding a delay time of control rod (CR) insertion initiation which was the most sensitive uncertainty to the results is also shown.
A plenty of plutonium is dealt in Plutonium Fuel Fabrication Facility and the facility is required to confine plutonium within a limited space such as glove box (GB) because plutonium is-emitter and causes an internal exposure. The MOX particles entrainment occurs and some of them are transiting to the outlet of GB without deposition to floor and wall. The entraining rate and the transiting rate are reported as Airborne Release Fraction (ARF) and Respirable Fraction (RF) in the literatures. However, no activities of model development and analytical approach have been found for ARF and RF. Thus, a feasibility study is done in this paper on the behavior of MOX particles in GB such as entraining and transiting. A modeling code has been developed by improving AQUA-SF code and the RF values for abnormal occurrences, such as free-fall spill, outflow and fire, have been analyzed and compared with those reported. This paper also shows the analytical results of the improved code together with the simulated experimental results. It is found that the calculated values are almost corresponded to those reported and that the improved code can estimate MOX particle behavior in GB well.
scite is a Brooklyn-based organization that helps researchers better discover and understand research articles through Smart Citations–citations that display the context of the citation and describe whether the article provides supporting or contrasting evidence. scite is used by students and researchers from around the world and is funded in part by the National Science Foundation and the National Institute on Drug Abuse of the National Institutes of Health.
hi@scite.ai
10624 S. Eastern Ave., Ste. A-614
Henderson, NV 89052, USA
Copyright © 2024 scite LLC. All rights reserved.
Made with 💙 for researchers
Part of the Research Solutions Family.