scholarly journals Evaluation of the Potential for Containment Bypass due to Steam Generator Tube Rupture in VVER-1000/V320 Reactor during Extended SBO sequence using SCDAP/RELAP5 code

2021 ◽  
Vol 9 (4) ◽  
pp. 9-15
Author(s):  
Van Thai Nguyen ◽  
Manh Long Doan ◽  
Chi Thanh Tran
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Severe Accident ◽  
Pressure Relief ◽  
Plant Behavior ◽  
Pressure Relief Valves ◽  
Station Blackout ◽  
Coolant System ◽  
Relap5 Code ◽  
Tube Failure

A severe accident-induced of a Steam Generator (SG) tube releases radioactivity from the Reactor Coolant System (RCS) into the SG secondary coolant system from where it may escape to the environment through the pressure relief valves and an environmental release in this manner is called “Containment Bypass”. This study aims to evaluate the potential for “Containment Bypass” in VVER/V320 reactor during extended Station Blackout (SBO) scenarios that challenge the tubes by primarily involving a natural circulation of superheated steam inside the piping loop and then induce creep rupture tube failure. Assessments are made of SCDAP/RELAP5 code capabilities for predicting the plant behavior during an SBO event and estimates are made of the uncertainties associated with the SCDAP/RELAP5 predictions for key fluid and components condition and for the SG tube failure margins. 

Natural Circulation and Creep Damage Analyses of Severe Accident for a Typical PWR Plant

2009 ◽  
Author(s):  
Osamu Kawabata ◽  
Masao Ogino
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Reactor Core ◽  
Creep Damage ◽  
Reactor Vessel ◽  
Reactor System ◽  
Severe Accident ◽  
Steam Generator Tube ◽  
Hot Gas ◽  
Core Meltdown

When the primary reactor system remain pressurized during core meltdown for a typical PWR plant, loop seals formed in the primary reactor system would lead to natural circulations in hot leg and steam generator. In this case, the hot gas released from the reactor core moves to a steam generator, and a steam generator tube would be failed with cumulative creep damage. From such phenomena, a high-pressure scenario during core meltdown may lead to large release of fission products to the environment. In the present study, natural circulation and creep damage in the primary reactor system accompanying the hot gas generation in the reactor core were discussed and the combining analysis with MELCOR and FLUENT codes were performed to examine the natural circulation behavior. For a typical 4 loop PWR plant, MELCOR code which can analyze for the severe accident progression was applied to the accident analyses from accident initiation to reactor vessel failure for the accident sequence of the main steam pipe break which is maintained at high pressure during core meltdown. In addition, using the CFD code FLUENT, fluid dynamics in the reactor vessel plenum, hot leg and steam generator of one loop were simulated with three-dimensional coordinates. And the hot gas natural circulation flow and the heat transfer to adjoining structures were analyzed using results provided by the MELCOR code as boundary conditions. The both ratios of the natural circulation flow calculated in the hot leg and the steam generator using MELCOR code and FLUENT code were obtained to be about 2 (two). And using analytical results of thermal hydraulic analysis with both codes, creep damage analysis at hottest temperature points of steam generator tube and hot leg were carried out. The results in both cases showed that a steam generator tube would be failed with creep rupture earlier than that of hot leg rupture.

CFD Predictions of Severe Accident Steam Generator Flows in a 1/7th Scale Pressurized Water Reactor

2002 ◽  
Author(s):  
Christopher Boyd ◽  
Kelly Hardesty
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Flow Phenomena ◽  
Experimental Values ◽  
High Degree ◽  
Side Flow

Computational Fluid Dynamics (CFD) is applied to steam generator inlet plenum mixing as part of a larger plan covering steam generator tube integrity. The technique is verified by comparing predicted results with severe accident natural circulation data [1] from a 1/7th scale Westinghouse facility. This exercise demonstrates that the technique can predict the natural circulation and mixing phenomena relevant to steam generator tube integrity issues. The model includes primary side flow paths for a single hot leg and steam generator. Qualitatively, the experimentally observed flow phenomena are predicted. The paths of the natural circulation flows and the relative flow proportions are correctly predicted. Quantitatively, comparisons are made with temperatures, mass flows, and other parameters. All predictions are generally within 10% of the experimental values. Overall, there is a high degree of confidence in the CFD technique for prediction of the relevant flow phenomena associated with this type of severe accident sequence.

Study on the Accidental Rupture of Hot Leg or Surge Line in SBO Accident

2006 ◽  
Author(s):  
Kun Zhang ◽  
Xuewu Cao
Keyword(s):  
Pressure Vessel ◽  
Severe Accident ◽  
Base Case ◽  
Management Guidelines ◽  
Accident Management ◽  
Lower Head ◽  
Station Blackout ◽  
Surge Line ◽  
Reactor Pressure ◽  
Relap5 Code

The postulated total station blackout accident (SBO) of PWR NPP with 600 MWe in China is analyzed as the base case using SCDAP/RELAP5 code. Then the hot leg or surge line are assumed to rupture before the lower head of Reactor Pressure Vessel (RPV) ruptures, and the progressions are analyzed in detail comparing with the base case. The results show that the accidental rupture of hot leg or surge line will greatly influence the progression of accident. The probability of hot leg or surge line rupture in intentional depressurization is also studied in this paper, which provides a suggestion to the development of Severe Accident Management Guidelines (SAMG).

Safety Analysis on Mitigation Overpressure ATWS in VVER-1000 NPP

2013 ◽  
Author(s):  
Ting Qi ◽  
Changjiang Yang
Keyword(s):  
Nuclear Power ◽  
Late Phase ◽  
Safety Analysis ◽  
Severe Accident ◽  
Injection System ◽  
Feed Water ◽  
System Pressure ◽  
Design Characteristics ◽  
Coolant System ◽  
Relap5 Code

Safety analysis on operating nuclear power plant (NPP) plays an important role on nuclear energy application, especially after the severe accident of Fukushima plants. This paper focuses on one of the operating NPPs in China, the TianWan NPP Unit1&2. TianWan Unit1&2 belong to VVER-1000 NPP type, which have special design characteristics and different safety migrating methods comparing with other domestic PWR NPPs in China. Calculations and analyses were made to give thermal hydraulics support to Level-1 PSA of this VVER type PWR on the anticipated transient without scram (ATWS) accidents. The calculation of loss of main feed water ATWS was carried out using the RELAP5 code to evaluate the intrinsic safety mainly impacted by fuel and moderator. The model also considers reactivity introduction caused by the change of boron concentration to find out the influence of the emergency boron injection system (JDH) on mitigating the coolant system over pressure. The paper gives out the success criteria of the safety valves of pressurizer (PRZ) with the critical moderator temperature coefficient (MTC) value according to the condition of coolant system pressure being under the pressure limit. It is indicated that the JDH system can play an important role on mitigating the over pressure of coolant system in the late phase of the transient.

scholarly journals Safety Analyses of VVER-1200/V491 reactor for longterm station blackout along with small LOCAs

2016 ◽  
Vol 6 (4) ◽  
pp. 8-17
Author(s):  
Thi Hoa Bui ◽  
Tan Hung Hoang ◽  
Minh Giang Hoang
Keyword(s):  
Steam Generator ◽  
Heat Removal ◽  
Safety Analysis ◽  
Severe Accident ◽  
Management Guideline ◽  
Accident Management ◽  
Station Blackout

Performance of  Passive Heat Removal through Steam Generator (PHRS-SG) of VVER-1200/V491 reactor presented in Safety Analysis Report for Ninh Thuan 1 shows that in case of long term station black out (SBO),  VVER-1200/V491 reactor can be cooldown and remained in safety mode at least 24 hours based on PHRS-SG performance. Anyway, long term station blackout along with small break in main coolant pipe of VVER-1200/V491 is assumed to be happening as an extension design condition that needs to be investigated. This study focuses on investigation of SBO along with different size of small break of LOCAs with expectation of finding the range of break size that the reactor is still kept in safety mode during 24 hours. During the investigation, some indicators for fuel damage such as the timing of HA1 actuation or mass of coolant inventory discharged are introduced as necessary information contributed to Severe Accident Management Guideline (SAMG).

RELAP5 Code Study of ROSA/LSTF Experiment on a PWR Station Blackout (TMLB’) Transient

2013 ◽  
Author(s):  
Takeshi Takeda ◽  
Hideo Nakamura
Keyword(s):  
Reverse Flow ◽  
Natural Circulation ◽  
Sensitivity Analyses ◽  
Test Analysis ◽  
Coolant Injection ◽  
Post Test ◽  
Station Blackout ◽  
Core Cooling ◽  
Secondary Side ◽  
Relap5 Code

RELAP5 code post-test analysis was performed on one of abnormal transient tests conducted with the ROSA/LSTF simulating a PWR station blackout (SBO) transient with the TMLB’ scenario in 1995. The RELAP5 code predicted well the overall trend of the major phenomena observed in the LSTF test, and indicated remaining problems in the predictions of reverse flow U-tubes in steam generator (SG) during long-term single-phase liquid natural circulation. Sensitivity analyses were done further to clarify effectiveness of depressurization of and coolant injection into SG secondary-side as accident management measures to maintain core cooling, based on the LSTF post-test analysis. SG secondary-side depressurization was initiated by fully opening the safety valve in one of two SGs with the incipience of core uncovery. Coolant injection was done into the secondary-side of the same SG at low pressures considering availability of fire engines. The SG depressurization with the coolant injection was found to well contribute to maintain core cooling by the actuation of accumulator system during a PWR SBO (TMLB’) transient.

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