Development of a phenomena identification and ranking table for thermal-hydraulic phenomena during a double-ended guillotine break LOCA in an SRS production reactor

Hanson, R.G.; Ortiz, M.G.; Bolander,; Wilson, G.E.

doi:10.2172/6528089

Cited by 3 publications

(3 citation statements)

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“…The code's applicability to these conditions has been ensured through an extensive code assessment, verification, and validation process. For LBLOCAs, these efforts were guided by three LBLOCA-specific phenomena identification and ranking tables (PIRTs) (Shaw, Larson, & Dimenna, 1988) (Young, Bajorek, Nissley, & Hochreiter, 1998) (Martin & O'Dell, 2005). Important phenomena were identified across all major phases of LBLOCA events-blowdown, refill, and reflood-and were used to inform the TRACE validation and development efforts.…”

Section: Tracementioning

confidence: 99%

Comparison Between Pin-by-Pin Subchannel and System Level Thermal Hydraulic Results for High Burnup Loss-of-Coolant Applications

Salko Jr.,

Wysocki,

Hizoum

et al. 2023

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This milestone report summarizes recent work to investigate higher fidelity modeling and simulation practices for large-break loss-of-coolant accident (LBLOCA) analysis in high-burnup pressurized water reactor (PWR) cores. Because of current industry interest in extending fuel cycle lengths ranging from 18 to 24-months, the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has been investigating the susceptibility of high-burnup core designs to fuel fragmentation, release, and dispersal (FFRD) during accident conditions such as LBLOCA. This work has prioritized developing and demonstrating a methodology for calculating core-wide susceptibility to FFRD and addressing uncertainties identified in the analysis and the US Nuclear Regulatory Commission (NRC) Research Information Letter on FFRD. Part of this investigation seeks to quantify differences between system-level thermal hydraulic behavior and higher fidelity subchannel modeling methods to identify potential safety concerns or opportunities to minimize FFRD susceptibility.To this end, the NEAMS subchannel code, CTF, is being used, along with the NRC system analysis code TRACE, for analysis of LBLOCA in a core containing high-burnup fuel. This project includes two thrusts: (1) improving on the existing TRACE model for a 4-loop PWR for LBLOCA so that the model is higher fidelity and consistent with current USNRC best practices, and (2) using CTF to perform pinresolved modeling of the core region of the reflood phase of a LBLOCA in a high-burnup PWR.Although CTF includes a wide range of closure models for modeling LOCA conditions, these models have not been assessed during recent CTF development activities, so the first step is to validate CTF using loss-of-coolant accident (LOCA) reflood experiments. A sensitivity study was also performed to identify high-impact models, which were then reviewed and corrected as needed. To strengthen the assessment, a code-to-code benchmark was performed against TRACE for these conditions. Both highand low-flooding rate tests from the FEBA reflood experiments were modeled using lumped and pinresolved CTF models, and the models were compared against lumped pin TRACE results. The lumped CTF and TRACE models demonstrated good agreement with experimental peak cladding temperatures (PCT) for all axial levels except near the top of the bundle. CTF pin-resolved PCT predictions were good for all thermocouple locations except for those facing the center subchannel, which were predicted to be about 40 K too low. A study on radial effects on the pin-resolved model was performed, and advanced CTF spacer grid droplet breakup and quenching models were tested to demonstrate the impact on better cooling the top of the bundle for better agreement with experimental data.Several modifications and improvements were made to the full-system TRACE model in this work, to better align the model with industry and NRC best practices and improve the accuracy of the LBLOCA predictions. This included converting to the best-practice 3D vessel comp...

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Section: Tracementioning

confidence: 99%

Comparison Between Pin-by-Pin Subchannel and System Level Thermal Hydraulic Results for High Burnup Loss-of-Coolant Applications

Salko Jr.,

Wysocki,

Hizoum

et al. 2023

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“…Uncertainty analysis can be a tool to help improve these activities to make current and future power plants more resilient to the potentially severe consequences of unmitigated NPP accidents. Preliminary phenomena identification and ranking tables for severe accidents have been established [12], but they are still not as detailed as for design basis accident analyses [13][14][15].…”

Section: Introductionmentioning

confidence: 99%

Uncertainty Study of the In-Vessel Phase of a Severe Accident in a Pressurized Water Reactor

Šadek

Grgić

Allison³

et al. 2022

Energies

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A comprehensive uncertainty analysis in the event of a severe accident in a two-loop pressurized water reactor is conducted using an uncertainty package integrated in the ASYST code. The plant model is based on the nuclear power plant (NPP) Krško, a Westinghouse-type power plant. The station blackout scenario with a small break loss of coolant accident is analyzed, and all processes of the in-vessel phase are covered. A best estimate plus uncertainty (BEPU) methodology with probabilistic propagation of input uncertainty is used. The uncertain parameters are selected based on their impact on the safety criteria, the operation of the NPP safety systems and to describe uncertainties in the initial and boundary conditions. The number of required calculations is determined by the Wilks formula from the desired percentile and confidence level, and the values of the uncertain parameters are randomly sampled according to appropriate distribution functions. Results showing the thermal hydraulic behaviour of the primary system and the propagation of core degradation are presented for 124 successful calculations, which is the minimum number of required calculations to estimate a 95/95 tolerance limit at the 3rd order of the Wilks formula application. A statistical analysis of the dispersion of results is performed afterwards. Calculation of the influence measures shows a strong correlation between the decay heat and the representative output quantities, which are the mass of hydrogen produced during the oxidation and the height of molten material in the lower head. As the decay heat increases, an increase in the production of hydrogen and the amount of molten material is clearly observed. The correlation is weak for other input uncertain parameters representing physical phenomena, initial and boundary conditions. The influence of the order of the Wilks formula is investigated and it is found that increasing the number of calculations does not significantly change the bounding values or the distribution of results for this particular application.

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“…In 1988, the Nuclear Regulatory Commission deployed the Code Scaling, Applicability, and Uncertainty evaluation methodology for emergency core cooling systems of power plants [7], which was demonstrated on a large break loss of coolant accident in a pressurized water reactor and published as a collection of papers [8][9][10][11][12]. They produced a phenomena identification and ranking table [8,13] according to the expected effect on the response variable peak clad temperature. Further screening from analytical sensitivity calculations [14] yielded a set of 28 inputs/parameters [9].…”

Section: Introductionmentioning

confidence: 99%

Closure Law Model Uncertainty Quantification

Strand

Kjølaas

Bergstrøm

et al. 2022

Int. J. UncertaintyQuantification

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The prediction uncertainty in simulators for industrial processes is due to uncertainties in the input variables and uncertainties in specification of the models, in particular the closure laws. In this work, the uncertainty in each closure law was modeled as a random variable and the parameters of its distribution were optimized to correctly quantify the uncertainty in predictions. We have developed two methods for optimization, based on the integrated quadratic distance and the energy score. The proposed methods were applied to the commercial multiphase flow simulator LedaFlow with the liquid volume fraction and pressure gradient as output variables. Two datasets were analyzed. Both describe two-phase gas-liquid flow, but are otherwise fundamentally different. One is gas-dominated stratified/annular flow and the other is liquiddominated slug flow. The closure law for the gas-wall friction factor is decisive for the gas-dominated predictions, and the estimated relative standard deviation is 4.5 % or 8.0 % depending on method. The liquid-dominated study showed that the liquid-wall friction factor and the slug bubble velocity are the closure laws with the greatest impact. Moreover, the estimated relative standard deviation in the liquid-wall friction factor is 5 %, and the deviation in the slug bubble velocity is 4 %. We used direct measurements of the slug bubble velocity to validate the estimated uncertainty.

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Development of a phenomena identification and ranking table for thermal-hydraulic phenomena during a double-ended guillotine break LOCA in an SRS production reactor

Cited by 3 publications

References 0 publications

Comparison Between Pin-by-Pin Subchannel and System Level Thermal Hydraulic Results for High Burnup Loss-of-Coolant Applications

Comparison Between Pin-by-Pin Subchannel and System Level Thermal Hydraulic Results for High Burnup Loss-of-Coolant Applications

Uncertainty Study of the In-Vessel Phase of a Severe Accident in a Pressurized Water Reactor

Closure Law Model Uncertainty Quantification

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