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Title: Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis

Abstract

In order to utilize reduced enrichment fuel, the three-element-core design for the Advanced Neutron Source has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. To assess the impact of changes in the core region configuration and the thermal-hydraulic steady-state conditions, the safety analysis has been updated. This report gives the safety margins for the loss-of-off-site power and pressure-boundary fault accidents based on the RELAP5 results. AU margins are greater for the three-element-core simulations than those calculated for the two-element core.

Authors:
; ; ;
Publication Date:
Research Org.:
Oak Ridge National Lab., TN (United States)
Sponsoring Org.:
USDOE Office of Energy Research, Washington, DC (United States)
OSTI Identifier:
491437
Report Number(s):
ORNL/TM-13077
ON: DE97006847; TRN: 97:012625
DOE Contract Number:
AC05-96OR22464
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: Feb 1996
Country of Publication:
United States
Language:
English
Subject:
07 ISOTOPE AND RADIATION SOURCE TECHNOLOGY; 22 NUCLEAR REACTOR TECHNOLOGY; NEUTRON SOURCE FACILITIES; REACTOR SAFETY; REACTOR ACCIDENTS; DESIGN; HEAT TRANSFER; HYDRAULICS; FUEL ELEMENTS

Citation Formats

Chen, N.C.J., Wendel, M.W., Yoder, G.L., and Harrington, R.M. Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis. United States: N. p., 1996. Web. doi:10.2172/491437.
Chen, N.C.J., Wendel, M.W., Yoder, G.L., & Harrington, R.M. Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis. United States. doi:10.2172/491437.
Chen, N.C.J., Wendel, M.W., Yoder, G.L., and Harrington, R.M. Thu . "Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis". United States. doi:10.2172/491437. https://www.osti.gov/servlets/purl/491437.
@article{osti_491437,
title = {Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis},
author = {Chen, N.C.J. and Wendel, M.W. and Yoder, G.L. and Harrington, R.M.},
abstractNote = {In order to utilize reduced enrichment fuel, the three-element-core design for the Advanced Neutron Source has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. To assess the impact of changes in the core region configuration and the thermal-hydraulic steady-state conditions, the safety analysis has been updated. This report gives the safety margins for the loss-of-off-site power and pressure-boundary fault accidents based on the RELAP5 results. AU margins are greater for the three-element-core simulations than those calculated for the two-element core.},
doi = {10.2172/491437},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Thu Feb 01 00:00:00 EST 1996},
month = {Thu Feb 01 00:00:00 EST 1996}
}

Technical Report:

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  • A summary of the methods and models used to perform neutronics analyses on the Advanced Neutron Source reactor three-element core design is presented. The applications of the neutral particle Monte Carlo code MCNP are detailed, as well as the expansion of the static role of MCNP to analysis of fuel cycle depletion calculations. Results to date of these applications are presented also. A summary of the calculations not yet performed is also given to provide a {open_quotes}to-do{close_quotes} list if the project is resurrected.
  • In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangersmore » are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components.« less
  • Two-dimensional discrete ordinates radiation transport calculations were performed for a model of the three-element core Advanced Neutron Source reactor design under normal operating conditions. The core consists of two concentric upper elements and a lower element radially centered in the annulus between the upper elements. The initial radiation transport calculations were performed with the DORT two-dimensional discrete ordinates radiation transport code using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub 6} quadrature, and a P{sub 1} Legendre polynomial expansion of the cross sections to determine the fission neutron source distribution in the core fuel elements. These calculations were limited to neutronmore » groups only. The final radiation transport calculations, also performed with DORT using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub l0} quadrature, and a P{sub 3} Legendre polynomial expansion of the cross sections, produced neutron and gamma-ray fluxes over the full extent of the geometry model. Responses (or activities) at various locations in the model were then obtained by folding the appropriate response functions with the fluxes at those locations. Some comparisons were made with VENTURE-calculated (diffusion theory) 20-group neutron fluxes that were summed into four broad groups. Tne results were in reasonably good agreement when the effects of photoneutrons were not included, thus verifying the physics model upon which the shielding model was based. Photoneutrons increased the fast-neutron flux levels deep within the D{sub 2}0 several orders of magnitude. Results are presented as tables of activity values for selected radial and axial traverses, plots of the radial and axial traverse data, and activity contours superimposed on the calculational geometry model.« less
  • Calculations of several important neutronic parameters have been performed for ten different three-element configurations considered for the Advanced Neutron Source (ANS) Reactor. Six of these configurations (labeled ST, SB, MT, MB, LT, and LB) are there result of the permutations of the same three elements. Two configurations (ST- MOD and SB-MOD) have the same element configuration as their base core design (ST and SB) but have slightly different element dimensions, and two configurations (ST-OL1 and ST-OL2) have two overlapping elements to increase the neutron fluxes in the reflector. For each configuration, in addition to the conceptual two-element design, fuel-cycle calculationsmore » were performed with calculations required to obtain unperturbed fluxes. The element power densities, peak thermal neutron flux as a function of position throughout the cycle, fast flux, fast-to-thermal flux ratios, irradiation and production region fluxes, and control rod worth curves were determined. The effective multiplication factor for each fuel element criticality. A comparison shows that the ST core configurations have the best overall performance, and the fully overlapping core configuration ST-OL2 has the best performance by a large margin. Therefore, on the basis of the neutronics results, the fully overlapping configuration is recommended for further consideration in using a three-element ANS reactor core. Other considerations such as thermal-hydraulics, safety, and engineering that are not directly related to the core neutronic performance must be weighed before a final design is chosen.« less
  • A reactor physics analysis was performed for the Advanced Neutron Source reactor with a three-element core configuration. The analysis was performed with a two-dimensional r-z 20-energy-group finite-difference diffusion theory model of the 17-d fuel cycle. The model included equivalent r-z geometry representations of the central control rods, the irradiation and production targets, and reflector components. Calculated quantities include fuel cycle parameters, fuel element power distributions, unperturbed neutron fluxes in the reflector and target regions, reactivity perturbations, and neutron kinetics parameters.