Severe accident natural circulation studies at the INEL

Bayless, Paul D.; Brownson, D.A.; Dobbe, C.A.; Jones, Kenneth R.; O’Brien, James E.; Pafford, D.J.; Schlenker, L.D.; Tung, V.X.

doi:10.2172/10120025

Cited by 14 publications

(9 citation statements)

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“…As shown in Figure 1, the present methodology to evaluate the temperature-induced RCS boundary creep rupture failure probability is to combine the existing methods for the creep rupture analysis [1,[8][9][10] with random data coming from the crack depth and the length adopted by getting insight from the probabilistic fracture mechanics. For specifically, the present approach consists of two parts: (1) thermal hydraulic …”

Section: Analysis Methodologymentioning

confidence: 99%

“…The pressure of the RCS remains at the set-point of pressurizer safety valves (PSVs) and the RCS inventory gradually reduces due to discharge through periodic opening of PSVs. It had been observed and proven by experimental and theoretical studies that, if this condition continues, the RCS is filled with saturated and/or superheated steam and non-condensable gases, and a severe accident natural circulation flow would be established [1]. The severe accident natural circulation flow is important because it transfers energy from the reactor core to other parts of the RCS.…”

Section: Introductionmentioning

confidence: 99%

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Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

Bang¹,

Jung²,

Lee³

et al. 2012

Journal of Nuclear Science and Technology

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A severe accident has inherently significant uncertainties due to the complex phenomena and wide range of conditions. Because of its high temperature and pressure, performing experimental validation and practical application are extremely difficult. With these difficulties, there has been few experimental researches performed and there is no plant-specific experimental data. Instead, computer codes have been developed to simulate the accident and have been used conservative assumptions and margins. This study is an effort to reduce the uncertainty in the probabilistic safety assessment and produce a realistic and physical-based failure probability. The methodology was developed and applied to the OPR1000. The creep rupture failure probabilities of reactor coolant system (RCS) components were evaluated under a station blackout severe accident with all powers lost and no recovery of steam generator auxiliary feedwater. The MELCOR 1.8.6 code was used to obtain the plant-specific pressure and temperature history of each part of the RCS and the creep rupture failure times were calculated by the rate-dependent creep rupture model with the plant-specific data.

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Section: Analysis Methodologymentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

Bang¹,

Jung²,

Lee³

et al. 2012

Journal of Nuclear Science and Technology

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“…Finally, the hot leg creep rupture calculation was configured in all three codes to use properties corresponding to the hot leg piping, rather than the hot leg nozzle safe end. Based on previous work (Bayless et al, 1995;EPRI, 2002), the last two assumptions will tend to substantially delay the timing of hot leg creep rupture relative to surge line and tube failure. Thus, the predicted results should only be used for comparing model predictions and should not be considered to reflect what would actually fail first.…”

Section: Modeling Of Natural Circulationmentioning

confidence: 96%

Severe accident analysis of a PWR station blackout with the MELCOR, MAAP4 and SCDAP/RELAP5 codes

Vierow¹,

Liao²,

Johnson³

et al. 2004

Nuclear Engineering and Design

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Based on knowledge obtained from experimental programs, considerable progress has been made in severe accident code development. The three leading severe accident codes used in the U.S., MELCOR, MAAP4 and SCDAP/RELAP5, are compared herein as part of an evaluation of the relative state of severe accident modeling in each of the codes.The MELCOR code is evolving from a probabilistic risk assessment tool to a best-estimate severe accident system analysis code. Some advantages that MELCOR has are the capability to evaluate containment behavior and the source term to the environment, and the great modeling flexibility that the control volume approach and control functions afford. The MAAP4 code was developed to perform fast-running full simulations of severe accidents. Due in part to the simplified form of the conservation equations and the coarser discretization of the reactor systems, MAAP4 has calculation times far shorter than those of the other codes while producing credible results. The SCDAP/RELAP5 code contains more mechanistic physics models than the other codes for both severe accident and thermal-hydraulic phenomena and has undergone extensive validation against plant and experimental data.The codes' overall attributes, relevant physics models and calculation results are compared herein. A hypothetical TMLB' scenario (station blackout with no recovery of auxiliary feedwater) at a 4-loop pressurized water reactor (PWR) was calculated by all three codes. Detailed plots showed that, despite considerable differences in the codes themselves, the calculation results of the codes are very similar in terms of thermal-hydraulic and core degradation response. There are minor discrepancies in various timings of phenomena, which are within the uncertainties of the code numerical computation and the physics models. The thermal challenge to the primary loop radionuclide barrier, such as the steam generator tube heat structures, is also compared among the three codes.

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“…Moreover, the base nodalization scheme on which IG-PWR is based has been thoroughly reviewed and validated during past INL projects. 68 Concerning the applicability of the RELAP5-3D code in simulating LB-LOCA and SBO events, other references can be found in the open literature, e.g., Refs. 69 and 70.…”

Section: Iiic Thermal-hydraulic Analysis Tool and Modelsmentioning

confidence: 99%

Risk-Informed Safety Analysis for Accident Tolerant Fuels

Parisi

Mandelli

et al. 2020

Nuclear Science and Engineering

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Accident tolerant fuels (ATFs) are being tested by different nuclear vendors and research organizations, and their introduction into the U.S light water reactor fleet is planned for the second half of the 2020s. Under the framework of the U.S. Department of Energy Light Water Reactor Sustainability (LWRS) Program, as part of the LWRS Risk-Informed Systems Analysis Pathway, research is being conducted at the Idaho National Laboratory (INL) to develop tools and methods that can help the industry to quantify the benefits from adopting ATF technology. In this paper we describe the developed risk-informed methodology including the safety analysis code improvements, and we present some results for selected accident scenarios. The developed methodology combines the INL state-of-the-art deterministic Best Estimate code RELAP5-3D and the probabilistic risk analysis tools RAVEN and SAPHIRE. The analyses are performed on a three-loop pressurized water reactor, simulating station blackout and large-break loss-of-coolant accidents and considering near-term ATFs or iron-chromium-aluminum and chromium-coated clads. Finally, we show how, applying our methodology, the new core damage frequency (CDF) can be assessed. The results indicate that the main benefit in introducing near-term ATFs is a significant reduction in hydrogen production during accident conditions. No significant CDF reduction was found.

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Severe accident natural circulation studies at the INEL

Cited by 14 publications

References 0 publications

Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

Severe accident analysis of a PWR station blackout with the MELCOR, MAAP4 and SCDAP/RELAP5 codes

Risk-Informed Safety Analysis for Accident Tolerant Fuels

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