skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Implementation of nodal equivalence parameters in DIF3D-VARIANT for core analysis of prismatic Very High Temperature Reactor (VHTR).

Abstract

The VARIANT module of the DIF3D code has been upgraded to utilize surface-dependent discontinuity factors. The performance of the new capability is verified using two-dimensional core cases with control rods in reflector and fuel blocks. Cross sections for VHTR components were generated using the DRAGON and HELIOS codes. For rodded block cross sections, the DRAGON calculations used a single-block model or the multi-block models combined with MCNP4C flux solutions, whereas the HELIOS calculations utilized multi-block models. Results from core calculations indicate that multiplication factor, block power, and control rod worth are significantly improved by using surface-dependent discontinuity factors.

Authors:
; ; ; ;
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
NE
OSTI Identifier:
925340
Report Number(s):
ANL-GENIV-092
TRN: US0803043
DOE Contract Number:
DE-AC02-06CH11357
Resource Type:
Technical Report
Country of Publication:
United States
Language:
ENGLISH
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; CONTROL ELEMENTS; CONTROL ROD WORTHS; CROSS SECTIONS; IMPLEMENTATION; MULTIPLICATION FACTORS; PERFORMANCE

Citation Formats

Lee, C. H., Joo, H. K., Yang, W. S., Taiwo, T. A., and Nuclear Engineering Division. Implementation of nodal equivalence parameters in DIF3D-VARIANT for core analysis of prismatic Very High Temperature Reactor (VHTR).. United States: N. p., 2007. Web. doi:10.2172/925340.
Lee, C. H., Joo, H. K., Yang, W. S., Taiwo, T. A., & Nuclear Engineering Division. Implementation of nodal equivalence parameters in DIF3D-VARIANT for core analysis of prismatic Very High Temperature Reactor (VHTR).. United States. doi:10.2172/925340.
Lee, C. H., Joo, H. K., Yang, W. S., Taiwo, T. A., and Nuclear Engineering Division. Thu . "Implementation of nodal equivalence parameters in DIF3D-VARIANT for core analysis of prismatic Very High Temperature Reactor (VHTR).". United States. doi:10.2172/925340. https://www.osti.gov/servlets/purl/925340.
@article{osti_925340,
title = {Implementation of nodal equivalence parameters in DIF3D-VARIANT for core analysis of prismatic Very High Temperature Reactor (VHTR).},
author = {Lee, C. H. and Joo, H. K. and Yang, W. S. and Taiwo, T. A. and Nuclear Engineering Division},
abstractNote = {The VARIANT module of the DIF3D code has been upgraded to utilize surface-dependent discontinuity factors. The performance of the new capability is verified using two-dimensional core cases with control rods in reflector and fuel blocks. Cross sections for VHTR components were generated using the DRAGON and HELIOS codes. For rodded block cross sections, the DRAGON calculations used a single-block model or the multi-block models combined with MCNP4C flux solutions, whereas the HELIOS calculations utilized multi-block models. Results from core calculations indicate that multiplication factor, block power, and control rod worth are significantly improved by using surface-dependent discontinuity factors.},
doi = {10.2172/925340},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Thu Mar 15 00:00:00 EDT 2007},
month = {Thu Mar 15 00:00:00 EDT 2007}
}

Technical Report:

Save / Share:
  • Enhancements have been made to the REBUS-3/DIF3D code suite to facilitate its use for the design and analysis of prismatic Very High Temperature Reactors (VHTRs). A new cross section structure, using table-lookup, has been incorporated to account for cross section changes with burnup and fuel and moderator temperatures. For representing these cross section dependencies, three new modules have been developed using FORTRAN 90/95 object-oriented data structures and implemented within the REBUS-3 code system. These modules provide a cross section storage procedure, construct microscopic cross section data for all isotopes, and contain a single block of banded scattering data for efficientmore » data management. Fission products other than I, Xe, Pm, and Sm, can be merged into a single lumped fission product to save storage space, memory, and computing time without sacrificing the REBUS-3 solution accuracy. A simple thermal-hydraulic (thermal-fluid) feedback model has been developed for prismatic VHTR cores and implemented in REBUS-3 for temperature feedback calculations. Axial conduction was neglected in the formulation because of its small magnitude compared to radial (planar) conduction. With the simple model, the average fuel and graphite temperatures are accurately estimated compared to reference STAR-CD results. The feedback module is currently operational for the non-equilibrium fuel cycle analysis option of REBUS-3. Future work should include the extension of this capability to the equilibrium cycle option of the code and additional verification of the feedback module. For the simulation of control rods in VHTR cores, macroscopic cross section deviations (deltas) have been defined to account for the effect of control rod insertion. The REBUS-3 code has been modified to use the appropriately revised cross sections when control rods are inserted in a calculation node. In order to represent asymmetric core blocks (e.g., fuel blocks or reflector blocks containing asymmetric absorber rods), surface-dependent discontinuity factors based on nodal equivalence theory have been introduced into the nodal diffusion theory option of the DIF3D code (DIF3D-nodal) to improve modeling accuracy. Additionally, the discontinuity factors based on the Simplified Equivalence Theory (SET) have been incorporated as an alternative and may be employed for both the DIF3D-nodal and DIF3D-VARIANT (nodal transport) solution options. Two- and three-dimensional core calculations have been performed using the routines developed and modified in this work, along with cross sections generated from single fuel block and one-dimensional or two-dimensional fuel-reflector model. Generally, REBUS-3/DIF3D results for the core multiplication factor and power distribution are found to be in good agreement with reference results (generated with MCNP continuous energy calculations) particularly when discontinuity factors are applied. The DIF3D-VARIANT option was found to provide a more accurate solution in its diffusion approximation than the DIF3D-nodal option. Control rod worths can be estimated with acceptably small errors compared to MCNP results. However, estimation of the core power tilt needs to be improved by introducing the surface-dependent discontinuity factor capability in DIF3D-VARIANT.« less
  • No abstract prepared.
  • The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs themore » same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).« less
  • The potential impact of nuclear data uncertainties on a number of performance parameters (core and fuel cycle) of the prismatic block-type Very High Temperature Reactor (VHTR) has been evaluated and results are presented in this report. An uncertainty analysis has been performed, based on sensitivity theory, which underlines what cross-sections, what energy range and what isotopes are responsible for the most significant uncertainties. In order to give guidelines on priorities for new evaluations or validation experiments, required accuracies on specific nuclear data have been derived, accounting for target accuracies on major design parameters. Results of an extensive analysis indicate onlymore » a limited number of relevant parameters do not meet the target accuracies assumed in this work; this does not imply that the existing nuclear cross-section data cannot be used for the feasibility and pre-conceptual assessments of the VHTR. However, the results obtained depend on the uncertainty data used, and it is suggested to focus some future evaluation work on the production of consistent, as far as possible complete and user oriented covariance data.« less
  • Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculationsmore » were compared with preliminary temperature limits derived from the ASME pressure vessel code.« less