scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

Investigation on Fission Products Release Mitigated by In-Containment Relief Valve Under SGTR Accident

2018 ◽  
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Sung Joong Kim
Keyword(s):  
Steam Generator ◽  
Free Space ◽  
Power Plants ◽  
Fission Products ◽  
Base Case ◽  
Relief Valve ◽  
Flow Area ◽  
Steam Generator Tube ◽  
Radioactive Nuclides ◽  
Safety Features

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in Nuclear Power Plants (NPPs). However, in a bypass scenario of Steam Generator Tube Rupture (SGTR), radioactive nuclides are released to environment even if the containment is not ruptured. The radioactive nuclides are transported from primary to secondary systems through a broken steam generator tube during SGTR accident. Accordingly, the radioactive nuclides of the secondary system can be released to the environment through Main Steam Safety Valve (MSSV) or Atmospheric Dump Valve (ADV). Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise In-Containment Relief Valve (ICRV) from steam generator to the free space in the containment building of the Optimized Power Reactor 1000 MWe (OPR1000). This study focuses on the conceptual development of the mitigation strategy and MELCOR code was used for the numerical simulation. The MELCOR input model of OPR1000 consists of 58 control volumes and 161 flow paths. Safety features such as Pressurizer Safety Relief Valve (PSRV), Safety Injection Tanks (SITs), and MSSV were modeled in the MELCOR model. To initiate the SGTR scenario, a flow path between secondary and primary sides of Steam Generator (SG) was modeled with a flow area of 4.49 × 10−4 m2. The safety features were assumed that a few passive systems such as PSRV, MSSV, and SIT, were available. Under this condition, the ICRV connecting the SG and the free space in the containment such as dome and Reactor Drain Tank (RDT) were modeled. Specifications of the ICRV such as length, flow area, and valve opening condition were assumed to similar to those of the MSSV. Using these paths, three cases were considered; a base case, a case of steam release to the containment dome (CNMT case), and a case of release to the RDT (RDT case). Simulation results show that in the base case released radionuclides to the environment. In the other cases, the radioactive nuclides were not released to the environment although the containment pressure increased more than the base case, which is lack of the ICRV. As a result, the ICRV prevents the radionuclides release to the environment during SGTR accidents. Further studies are needed to incorporate practical valve inputs, reactor type, and safety features to gain more feasibility.

Sensitivity Analyses for a PWR SCC Case Using the Probabilistic Loss-of-Coolant-Accident (Pro-LOCA) Software

2016 ◽  
Author(s):  
Bruce A. Young ◽  
Sang-Min Lee ◽  
Paul M. Scott
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
The United States ◽  
Design Criterion ◽  
Mitigation Strategies ◽  
Sensitivity Analyses ◽  
Base Case ◽  
Piping System ◽  
Loss Of Coolant Accident ◽  
Piping Systems

As a means of demonstrating compliance with the United States Code of Federal Regulations 10CFR50 Appendix A, General Design Criterion 4 (GDC-4) requirement that primary piping systems for nuclear power plants exhibit an extremely low probability of rupture, probabilistic fracture mechanics (PFM) software has become increasingly popular. One of these PFM codes for nuclear piping is Pro-LOCA which has been under development over the last decade. Currently, Pro-LOCA is being enhanced under an international cooperative program entitled PARTRIDGE-II (Probabilistic Analysis as a Regulatory Tool for Risk-Informed Decision GuidancE - Phase II). This paper focuses on the use of a pre-defined set of base-case inputs along with prescribed variation in some of those inputs to determine a comparative set of sensitivity analyses results. The benchmarking case was a circumferential Primary Water Stress Corrosion Crack (PWSCC) in a typical PWR primary piping system. The effects of normal operating loads, temperature, leak detection, inspection frequency and quality, and mitigation strategies on the rupture probability were studied. The results of this study will be compared to the results of other PFM codes using the same base-case and variations in inputs. This study was conducted using Pro-LOCA version 4.1.9.

scholarly journals Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

2021 ◽  
Vol 13 (14) ◽  
pp. 7964
Author(s):  
Alain Flores y Flores ◽  
Danilo Ferretto ◽  
Tereza Marková ◽  
Guido Mazzini
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Accident Simulation ◽  
Experimental Facilities ◽  
The Difference

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

A Modeling Method of the Branch Probability of the Temperature-Induced SGTR Phenomenon

2017 ◽  
Author(s):  
Zibin Liu ◽  
Dingqing Guo ◽  
Bing Zhang ◽  
Jinkai Wang
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Modeling Method ◽  
Radioactive Material ◽  
Severe Accident ◽  
Steam Generator Tube ◽  
Modeling Methods ◽  
Radioactive Release ◽  
Level 2

The phenomenon of temperature-induced steam generator tube rupture (TI-SGTR) is a typical phenomenon in the severe accident process of nuclear power plants. The occurrence of the phenomenon may result in the radioactive material bypass the containment, causing a large radioactive release. This paper investigates modeling methods of the phenomenon of temperature-induced SGTR in level 2 PSA and presents an optimizing modeling method to calculate the probability of branching probability of TI-SGTR, aiming at improving the rationality and veracity of level 2 PSA.

Preliminary Study of Influence of Severe Accident Mitigation Countermeasures to In-Vessel Release of Fisson Products

2010 ◽  
Author(s):  
Gaofeng Huang ◽  
Xuewu Cao ◽  
Jingxi Li
Keyword(s):  
Nuclear Power ◽  
Fission Products ◽  
Early Time ◽  
Severe Accident ◽  
Base Case ◽  
Primary Loop ◽  
Radiological Consequence ◽  
Secondary Loop ◽  
Station Blackout ◽  
Preliminary Study

During the severe accident in a nuclear power plant, large amounts of fission products release with accident progression, which includes in-vessel release and ex-vessel release. Mitigation of release of fission products is the need of alleviating radiological consequence in severe accident. Mitigation countermeasures to in-vessel release of fission products are studied, including feed-bleed in primary loop, feed-bleed in secondary loop and cooling of ex-vessel. Representative high pressure melt accident of station blackout is chosen, and different entry condition of countermeasures is assumed. The results show that: (1) Feed-bleed in primary loop is an effective countermeasure to mitigate in-vessel release of fission products. With early time to implement the countermeasure, in-vessel release fraction of fission products is low. (2) Feed-bleed in secondary loop is also an effective countermeasure to mitigate in-vessel release of fission products. Low in-vessel release fraction of fission products is produced with early time of countermeasure implemented. (3) Cooling of ex-vessel is not an effective countermeasure to control in-vessel release of fission products, the in-vessel release fraction in this case is almost equal to base case that uses none countermeasure.

Sensitivity Analysis of RCS Depressurization Strategy Under a Postulated SGTR Accident in OPR1000

2018 ◽  
Author(s):  
Wonjun Choi ◽  
Taeseok Kim ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Sung Joong Kim
Keyword(s):  
Sensitivity Analysis ◽  
Fission Product ◽  
Steam Generator ◽  
Flow Rate ◽  
Sensitivity Study ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Feed Water ◽  
Relief Valve ◽  
Product Release

Steam generator tube rupture (SGTR) accident is one of important accident that has high probability of resulting in severe accidents. As a bypass scenario, fission product can be directly released to the environment during the SGTR accident. Thus, the severe accident by SGTR should be carefully managed by severe accident management guidance (SAMG). In Korea, SAMG for optimized power reactor 1000 (OPR1000) has been developed in 1999 and used to mitigate the severe accident of OPR1000 with seven mitigation strategies. Among the mitigation strategies, ‘Depressurization of reactor coolant system (RCS)’ is one of the most powerful strategies to reduce direct release of the fission product. To reduce the RCS pressure, indirect depressurization using steam generator is generally recommended. However, depending on the RCS condition, the indirect depressurization can be ineffective to reduce the RCS pressure. In this case, direct depressurization using pilot operated relief valve (PORV) should be performed as a second plan. From this point of view, sensitivity study of RCS depressurization was performed to investigate priority of depressurization in this study. The severe accident scenario initiated by SGTR accident was selected from probabilistic safety assessment (PSA) level 1 report and simulated using MELCOR 2.1. For the mitigation strategy, various timing of depressurization, the number of opening valves and flow rate of feed water were applied to simulate the possible depressurization strategies during the severe accident. The MELCOR code simulation shows that if depressurization was performed at 30 minutes after SAMG entrance, the direct depressurization was more efficient to reduce the RCS pressure and the fission product release. Therefore, it was recommended to use direct depressurization rather than indirect depressurization in certain time. The sensitivity of flow rate of feed water and different number of opening valves were insignificant for progress of the accident and fission product release. In conclusion, operators should select the way of depressurization to reduce the RCS pressure and the fission product release during the SGTR accident, considering the condition of the plant such as accident progress and availability of safety features. To suggest more proper information for depressurization, more sensitivity analysis and detailed thermal-hydraulic analysis should be performed for the future work.

scholarly journals Safety assessments and severe accidents, impact of external events on nuclear power plants and on mitigation strategies

2020 ◽  
Vol 6 ◽  
pp. 39
Author(s):  
Jean-Pierre Van Dorsselaere ◽  
Ahmed Bentaib ◽  
Thierry Albiol ◽  
Florian Fichot ◽  
Alexei Miassoedov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Severe Accidents ◽  
Accident Management ◽  
External Events ◽  
Safety Assessments ◽  
The University

The Fukushima-Daiichi accidents in 2011 underlined the importance of severe accident management (SAM), including external events, in nuclear power plants (NPP) and the need of implementing efficient mitigation strategies. To this end, the Euratom work programmes for 2012 and 2013 was focused on nuclear safety, in particular on the management of a possible severe accident at the European level. Relying upon the outcomes of the successful Euratom SARNET and SARNET2 projects, new projects were launched addressing the highest priority issues, aimed at reducing the uncertainties still affecting the main phenomena. Among them, PASSAM and IVMR project led by IRSN, ALISA and SAFEST projects led by KIT, CESAM led by GRS and sCO2-HeRO lead by the University of Duisburg-Essen. The aim of the present paper is to give an overview on the main outcomes of these projects.

Kinetic Study of Containment Retention Factor (CRF) for Large Dry Containment Under Radiation Load

2014 ◽  
Author(s):  
Khurram Mehboob ◽  
Kwangheon Park ◽  
Rehan Khan ◽  
Majid Ali ◽  
Raheel Ahmed
Keyword(s):  
Kinetic Study ◽  
Nuclear Power ◽  
Power Plants ◽  
Heat Removal ◽  
Fission Products ◽  
Retention Factor ◽  
Severe Accident ◽  
Spray System ◽  
Filtration System ◽  
Kinetic Dependency

The Nuclear Power Plants (NPPs) have been built on the concept of Defense in depth. The severe accident causes the failure of fission product barriers and let the fission products to escape into environment. The containment is the last barrier to the fission products. Thus, the containment is installed with engineering safety features (ESFs) i.e. spray system, heat removal system, recirculation filtration system; containment filtered venting system (CFVS), and containment exhaust filtration system. In this work, kinetic study of the containment retention factor (CRF) has been carried out for a large dry PWR containment considering 1000 MWe PWR. The computational modeling and simulation have been carried out by developing a kinetic code in MATLAB, which uses the fractions of activity airborne into the containment after the accident. The Kinetic dependency of CRF on containment filtration systems, spray system with caustic and boric acid spray has been carried out. For noble gases, iodine and aerosols, the CRF increases with the increase in exhaust rate. While, CRF for iodine first increases then start reducing with containment spray flow rate. The Kinetic dependency of CRF has also been studied for boric and caustic spray.

A Demonstration of Practical Elimination of Early or Large Fission Product Release for the UK ABWR Generic Design Assessment

2018 ◽  
Author(s):  
Ming Leang Ang ◽  
Nuh Mohamud ◽  
Hiromasa Chitose ◽  
Naoki Hirokawa ◽  
Ryusuke Kimura
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Severe Accident ◽  
Design Assessment ◽  
Safety Case ◽  
Safety Cases ◽  
The Uk ◽  
Generic Design

The demand for continuous improvement in safety of nuclear power plants has led to an international expectation that early or large releases of fission products as a result of severe accidents be practically eliminated for new reactor designs. The UK Department for Business, Energy and Industrial Strategy has recently published the Seventh UK Report which provided confirmation of UK demonstrating compliance with the obligations of the Convention on Nuclear Safety [Ref-1]. Relating to Article 14 on the Assessment and Verification of Safety, the UK nuclear Regulator (Office for Nuclear Regulation (ONR)) has the expectation of the demonstration of ‘practical elimination’ of potential severe accident states be included in the safety cases for new nuclear power plants [Ref-2]. In order to achieve this, the safety case should show either that it is physically impossible for the accident states to occur or the states can be considered to be extremely unlikely with a high degree of confidence by design provisions [Ref-3] [Ref-4] [Ref-5]. A demonstration framework was developed and applied successfully in the UK ABWR Generic Design Assessment (GDA) Pre-Construction Safety Report (PCSR) which was submitted to the ONR in August 2017 [Ref-6]. A summary of this demonstration is provided in this paper.

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