Multicode comparison of selected source-term computer codes

Hermann, O.W.; Parks, C.V.; Renier, J.P.; Roddy, J.W.; Ashline, R.C.; Wilson, W.B.; LaBauve, R.J.

doi:10.2172/6225635

Cited by 4 publications

(3 citation statements)

References 3 publications

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“…The neutronic capabilities of MCNP have been evaluated using PBR related benchmarks and are found to have excellent agreement with experimental predictions [18]. The ORIGEN methodology has been verified for various situations and, in general, good agreement has been found between the predictions of ORIGEN as compared to other codes and experimental data [19], [20]. The primary reasons for disagreement can generally be attributed to inaccuracies in the power history that is used in a given calculation, uncertainties in the used spectral-averaged cross-sections, and also possible uncertainties in the initial composition of the fuel (e.g., enrichment) [21].…”

Section: Simultaneous Burnup and Enrichment Determinationmentioning

confidence: 76%

Computational investigation of on-line interrogation of pebble bed reactor fuel

Hawari

Chen

2005

IEEE Trans. Nucl. Sci.

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Pebble bed reactors are characterized by multipass fuel systems in which spherical fuel pebbles are circulated through the core until they reach a proposed burnup limit (80 000-100 000 MWD/MTU). For such reactors, the fuel is assayed on-line to ensure that the burnup limit is not breached. We considered assaying the fuel using an HPGe detector to perform passive gamma-ray spectrometry of fission products. Since neither fresh nor irradiated fuel is readily available, computer simulations were utilized to identify the radionuclides that can be used as burnup indicators, and to visualize the gamma-ray spectra at various levels of burnup. Specifically, we used the ORIGEN-MONTEBURNS-MCNP code system. This allowed the establishment of the burnup dependent one-group gas reactor cross-sections for the radionuclides of interest. Subsequently, ORIGEN was used to simulate in-core pebble depletion to establish the irradiated pebble isotopics. Finally, the codes MCNP and SYNTH were used to simulate the response of the HPGe gamma-ray spectrometer. The results show that absolute and relative indicators can be used on-line to determine unambiguously the enrichment and burnup on a pebble-by-pebble basis. The activity of Cs-137 or the activity ratio of Co-60/Cs-134 can be combined with the activity ratio of Np-239/I-132 to yield the enrichment and burnup information. To use the relative indicators, a relative efficiency calibration of the gamma-ray spectrometer can be performed using the La-140 gamma lines that are emitted by the irradiated pebble. I-132, Cs-134, Cs-137, La-140, and Np-239 are produced upon the irradiation of the fuel. Co-60 is produced by doping the fuel with a small amount ( 100 ppm) of Co-59. Using this approach, the uncertainty in burnup determination due to factors such as power history variation, detector efficiency calibration, and counting statistics is expected to remain in the range of 5% to 10%.

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Section: Simultaneous Burnup and Enrichment Determinationmentioning

confidence: 76%

Computational investigation of on-line interrogation of pebble bed reactor fuel

Hawari

Chen

2005

IEEE Trans. Nucl. Sci.

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“…ORIGEN computes the time-dependent concentrations and source terms (radiation, heat) for over 1600 nuclides that are simultaneously generated or depleted through interactions of materials with neutrons and radioactive decay. ORIGEN has been subjected to numerous verification checks through intercode comparisons and validation tests by comparing code predictions with measured radionuclide concentrations in irradiated LWR and CANDU (CANada Deuterium Uranium) fuel (Tait et al, 1989), (Hermann et al, 1981), (Tait et al, 1995). Similarly, one-and two-dimensional depletion simulations can be performed with SCALE (-, 2009).…”

Section: Reactor Modelsmentioning

confidence: 99%

Integration of facility modeling capabilities for nuclear nonproliferation analysis

García¹,

Burr²,

Coles³

et al. 2012

Progress in Nuclear Energy

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Developing automated methods for data collection and analysis that can facilitate nuclear nonproliferation assessment is an important research area with significant consequences for the effective global deployment of nuclear energy. Facility modeling that can integrate and interpret observations collected from monitored facilities in order to ascertain their functional details will be a critical element of these methods. Although improvements are continually sought, existing facility modeling tools can characterize all aspects of reactor operations and the majority of nuclear fuel cycle processing steps, and include algorithms for data processing and interpretation. Assessing nonproliferation status is challenging because observations can come from many sources, including local and remote sensors that monitor facility operations, as well as open sources that provide specific business information about the monitored facilities, and can be of many different types. Although many current facility models are capable of analyzing large amounts of information, they have not been integrated in an analyst-friendly manner. This paper addresses some of these facility modeling capabilities and illustrates how they could be integrated and utilized for nonproliferation analysis. The inverse problem of inferring facility conditions based on collected observations is described, along with a proposed architecture and computer framework for utilizing facility modeling tools. After considering a representative sampling of key facility modeling capabilities, the proposed integration framework is illustrated with several examples. Highlights: We consider integration of facility modeling into a tool for proliferation analysis.  We describe the inverse problem of inferring the facility based on observables.  We propose architecture for integrating facility models into a suitable framework.  We illustrate the proposed integration framework with several examples.

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“…The model and corresponding data (Tables 2 and 3) axe taken from ref. 7. Initial efforts involved an investigation of the accuracy of the 1-D criticality predictions for the BWR fuel element model shown in Fig.…”

Section: Toolsmentioning

confidence: 99%

Feasibility Assessment of Burnup Credit in the Criticality Analysis of Shipping Casks with Boiling Water Reactor Spent Fuel

Broadhead

1995

Nuclear Technology

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Multicode comparison of selected source-term computer codes

Cited by 4 publications

References 3 publications

Computational investigation of on-line interrogation of pebble bed reactor fuel

Computational investigation of on-line interrogation of pebble bed reactor fuel

Integration of facility modeling capabilities for nuclear nonproliferation analysis

Feasibility Assessment of Burnup Credit in the Criticality Analysis of Shipping Casks with Boiling Water Reactor Spent Fuel

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