Accident Management Actions in an Upper-Head Small-Break Loss-of-Coolant Accident with High-Pressure Safety Injection Failed

Queral, César; Gonzalez-Cadelo, J.; Varas, Gonzalo Jiménez; Villalba, Ernesto

doi:10.13182/nt11-a12507

Cited by 25 publications

(4 citation statements)

References 7 publications

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“…The reactor thermal-hydraulic response to breaks of this size as initiating event has shown large sensitivity to the initial and plant-specific conditions. A small difference in the break size and location, thermal power, cross-over leg relative elevation, or manual actions taken on the secondary side can lead to a different plant evolution, affecting, among others, safety system actuations driving the plant to a safe state [35,36]. Nevertheless, responses to relatively small break LOCAs share the following characteristics until the onset of core damage [34]:…”

Section: Preliminary Remarks On the Scenariomentioning

confidence: 99%

ASTEC–MAAP Comparison of a 2 Inch Cold Leg LOCA until RPV Failure

Blul

Brumm

Mascari

et al. 2018

Science and Technology of Nuclear Installations

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A 2 inch, cold-leg loss-of-coolant accident (LOCA) in a 900 MWe generic Western PWR was simulated using ASTEC 2.1.1 and MAAP 5.02. The progression of the accident predicted by the two codes up to the time of vessel failure is compared. It includes the primary system depressurization, accumulator discharge, core heat-up, hydrogen generation, core relocation to lower plenum, and lower head breach. The purpose of the code comparison exercise is to identify modelling differences between the two codes and the user choices affecting the results. The two codes predict similar primary system depressurization behaviour until the accumulation injection, confirming similar break flow and primary system thermal-hydraulic response calculations between the two codes. The choice of the accumulator gas expansion model, either isentropic or isothermal, affects the rate and total amount of coolant injected and thereby determines whether the core is quenched or overheated and attains a noncoolable geometry during reflooding. A sensitivity case was additionally simulated by each code to allow comparisons to be made with either accumulator gas expansion models. The two codes predict similar amount of in-vessel hydrogen generated and core quench status for a given accumulator gas expansion model. ASTEC predicts much larger initial core relocation to lower plenum leading to an earlier vessel failure time. MAAP predicts more gradual core relocation to lower plenum, prolonging the lower plenum debris bed heat-up and time to vessel failure. Beside the effect of the code user in conducting severe accident simulations, some discrepancies are found in the modelling approaches in each code. The biggest differences are found in the in-vessel melt progression and relocation into the lower plenum, which deserve further research to reduce the uncertainties.

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Section: Preliminary Remarks On the Scenariomentioning

confidence: 99%

ASTEC–MAAP Comparison of a 2 Inch Cold Leg LOCA until RPV Failure

Blul

Brumm

Mascari

et al. 2018

Science and Technology of Nuclear Installations

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“…This is the case of the inspections performed in the Davis Besse reactor and the South Texas Project Unit-I reactor. In Davis Besse NPP (2002), a wall thinning and degradation in the reactor PV upper head was discovered as a result of circumferential cracking of a control rod drive mechanism penetration nozzle [1,2]. In the South Texas Unit-I reactor, a small amount of residue was detected around two instrument-tube penetration nozzles located in the PV lower plenum [3].…”

Section: Introductionmentioning

confidence: 99%

Break location influence in pressure vessel SBLOCA scenarios

Lorduy¹,

Querol²,

Gallardo³

et al. 2021

Braz. J. Rad. Sci.

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The inspections performed in Davis Besse and in the South Texas Project Unit-I reactors pointed out safety issues regarding the structural integrity of the Pressure Vessel (PV). In these inspections, two anomalies were found: a wall thinning and degradation in the PV upper head of the Davis Besse reactor and a small amount of residue around of two instrument-tube penetration nozzles located in the PV lower plenum of the South Texas Project Unit-I reactor. The evolution of these defects could have resulted in Small Break Loss-Of-Coolant Accidents (SBLOCA) if they had not been detected in time. In this frame, the OECD/NEA considered the necessity to simulate these accidental sequences in the OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). This work is focused on simulating different hypothetical accidental scenarios in the PV using the thermal-hydraulic code TRACE5. These simulations allow studying the break localization influence in the transient and the effectiveness of the accident management (AM) actions considered mitigating the consequences of these hypothetical accidental scenarios.

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“…Como se observa, la pendiente de dicha relación es similar en ambos modelos obteniendo que para el mismo valor de la CET, ambos casos predicen una PCT menor que en el experimento. En general, los resultados que se han obtenido con el código TRACE5 son similares al experimento y no distan mucho de los resultados obtenidos por otros autores (Freixa y Manera, 2010) y (Queral et al, 2011). Por eso, se ha considerado una buena simulación de los resultados del experimento 6-1 con el modelo de LSTF.…”

Section: Transitorio Del Experimento 6-1unclassified

“…Además, a partir de la inyección de los acumuladores, la presión del sistema primario en el modelo escalado cae más bruscamente, mientras que la del modelo de LSTF se comporta muy similar al experimento. Otros autores que también han escalado el transitorio 6-1 a un modelo de planta real (Freixa y Manera, 2011), (Queral et al, 2013) han obtenido resultados similares en cuanto a la caída más brusca de la presión del sistema primario en el modelo escalado. En dicho trabajo se justificaba que dicha caída de presión podría ser debida a diferencias geométrica en la vasija.…”

Section: Transitorio Del Experimento 6-1 Con El Modelo Escaladounclassified

Análisis termohidráulico de la instalación LSTF/ROSA utilizando el código TRACE5. Aportaciones a la metodología de escalado.

Vives¹

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i misma fenomenología que ocurre a pequeña escala. Para ello los resultados de simulación también se han comparado con los resultados experimentales y con los obtenidos con el modelo de LSTF.Además, para profundizar más en la metodología de escalado, se ha utilizado un modelo de la instalación PKL con el que se ha reproducido el experimento counterpart, común a las instalaciones LSTF y PKL, que tienen diferentes condiciones de diseño y operación. Esto permite estudiar la fenomenología y aspectos relativos al escalado en instalaciones con diferentes condiciones de diseño y operación, así como obtener las relaciones de escalado entre ambas instalaciones.

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Accident Management Actions in an Upper-Head Small-Break Loss-of-Coolant Accident with High-Pressure Safety Injection Failed

Abstract: Since the Three Mile Island accident, an important focus of pressurized water reactor (PWR) transient analyses has been a small-break loss-of-coolant accident (SBLOCA

Cited by 25 publications

References 7 publications

ASTEC–MAAP Comparison of a 2 Inch Cold Leg LOCA until RPV Failure

ASTEC–MAAP Comparison of a 2 Inch Cold Leg LOCA until RPV Failure

Break location influence in pressure vessel SBLOCA scenarios

Análisis termohidráulico de la instalación LSTF/ROSA utilizando el código TRACE5. Aportaciones a la metodología de escalado.

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