Establishment of analysis methodology of RADTRAD for Maanshan PWR ATWS
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H.-T. Lin
Abstract
In this research, the analysis methodology for Maanshan PWR nuclear power plant (NPP) ATWS (Anticipated Transient Without Scram) is established by using RADTRAD code. This methodology can evaluate the dose at Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and control room for the Maanshan ATWS. In addition, there are four leakage pathways in the Maanshan ATWS that are containment leakage, containment low volume purge, steam generator leakage, and ESF leakage. Therefore, for these leakage pathways, the RADTRAD models are established. Finally, according to the analysis results, they are below the criteria of 10 CFR 100.11 for the EAB and LPZ and Standard Review Plan (SRP) for the control room. Additionally, the steam generator leakage gives the maximum dose in these leakage pathways.
Kurzfassung
In diesem Beitrag wird die Anwendung des Programms RADT-RAD für Ausbreitungsrechnungen für den Druckwasserreaktor Maanshan beim Ereignis ATWS beschrieben. Mit diesem Programm können Dosisberechnungen für verschiedene Räume und Zonen bestimmt werden, hier werden die Dosen für Exclusion Area Boundary (EAB), für Low Population Zone (LPZ) und den Kontrollraum ausgewertet. Darüber hinaus gibt es vier Leckagepfade im Maanshan ATWS, nämlich Containmentleckage, Containment Low Volume Purge, Dampferzeugerleckage und ESF-Leckage. Daher werden für diese Leckagepfade die RADTRAD-Modelle erstellt. Gemäß den Analyseergebnissen liegen sie schließlich unter den Kriterien von 10 CFR 100.11 für die EAB und LPZ und dem Standard Review Plan (SRP) für den Kontrollraum. Außerdem folgt aus den Rechnungen, dass die Dampferzeugerleckage zur maximale Dosis in diesen Leckagepfaden führt.
Abbreviations
- ATWS
Anticipated Transient Without Scram
- BR
Breathing Rate
- DCF
Dose Conversion Factor
- EAB
Exclusion Area Boundary
- LPZ
Low Population Zone
- MURPU
Measurement Uncertainty Recapture Power Uprate
- NPP
Nuclear Power Plant
- PRT
Pressurizer Relief Tank
- RCP
Reactor Coolant Pump
- RCS
Reactor Coolant System
- SNL
Sandia National Laboratories
- SRP
Standard Review Plan
References
1 U.S. NRC: RADTRAD: a simplified model for radionuclide transport and removal and dose estimation, NUREG/CR-6604, 1997Search in Google Scholar
2 Westinghouse Electric Corporation: Westinghouse Anticipated Transients Without Trip Analysis, WCAP-8330, 1974Search in Google Scholar
3 Taiwan Power Company: Maanshan Nuclear Power Station Final Safety Analysis Report (FSAR), 2013Search in Google Scholar
4 Taiwan Power Company: Maanshan Nuclear Power Station Training materials, 2006Search in Google Scholar
5 Westinghouse Electric Corporation: Anticipated Transients Without Scram for Westinghouse Plants, 1979Search in Google Scholar
6 U.S. NRC: 10 CFR 100.11, 1997Search in Google Scholar
7 U.S. NRC: Standard Review Plan, NUREG-0800, Revision 3, 2007Search in Google Scholar
8 Wang, C. D.;Wang, C. Y.; Dai, L. C.; Lin, H. T.; Hsu, S. S.; Chan, Y. K.; Yuann, Y. R.: The ATWS Event Integral Analysis Methodology for Maanshan Nuclear Power Plant, 2019Search in Google Scholar
9 U.S. NRC: Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors, Regulatory Guide 1.195, 2003Search in Google Scholar
10 Eckerman, K. F.;Wolbarst, A. B.; Richardson, A. C. B.: Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, Federal Guidance Report No. 11, Sept. 1988, DOI:10.2172/629423310.2172/6294233Search in Google Scholar
11 Eckerman, K. F.; Ryman, J. C.: External Exposure to Radionuclides in Air, Water, and Soil, Federal Guidance Report No. 12, Sept. 1993.Search in Google Scholar
12 Wang, C. D.: Radiological Dose Analysis Methodology for Maanshan PWR Plant during Loss of Coolant Accident, INER-K0075R, 2018Search in Google Scholar
© 2021 Walter de Gruyter GmbH, Berlin/Boston, Germany
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- Establishment of analysis methodology of RADTRAD for Maanshan PWR ATWS
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Articles in the same Issue
- Frontmatter
- The simulation research on the natural circulation operation characteristic of FNPP in rolling and inclined condition
- Establishment of analysis methodology of RADTRAD for Maanshan PWR ATWS
- Calendar of Events
- CFD simulation of subcooled boiling flow in PWR 5 ⨯ 5 rod bundle
- Effectiveness assessment of improvement measures in physical protection system monitoring center
- A study on accumulator analysis for the valve performance evaluation system of nuclear power plants
- Calculation studies of coated particles performance in sodium-cooled fast reactor
- Effect of using graphene/water based nanofluid on heat transfer in heat exchangers with rotating straight inner tube
- Radiation shielding properties of mortars containing heavyweight particles
- Improvement of the passive efficiency calibration of the segmented gamma scanner
- Investigation of level density parameter dependence for some 233U, 235U, 237U, 239U, 249Cf, 251Cf, 237Pu and 247Cm nuclei in neutron fission cross sections with the incident energy up to 20 MeV
- Establishment of analysis methodology for ionizing mattresses using RESRAD-BUILD code
- Editorial
